WorldWideScience

Sample records for accident severity

  1. Management of severe accidents

    International Nuclear Information System (INIS)

    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)

  2. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    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. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

    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)

  4. 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.

  5. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    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

  6. 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

  7. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

    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 Research Program plan update

    International Nuclear Information System (INIS)

    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

  9. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

    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.

  10. 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.

  11. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

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

    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...... 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-state power following initial power excursion was in most cases approximately 20% of the nominal reactor power, according...

  12. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    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

  13. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

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

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

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

  15. Deterministic analyses of severe accident issues

    International Nuclear Information System (INIS)

    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

  16. Conclusions on severe accident research priorities

    International Nuclear Information System (INIS)

    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

  17. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    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

  18. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  19. 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).

  20. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

    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)

  1. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

    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

  2. Severe accident risks from external events

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

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

  3. Severe accident management concept for LWRS

    International Nuclear Information System (INIS)

    Although the advanced built-in engineered safety features and the highly trained personnel have led to extremely low probabilities of core melt accidents, there is a common understanding that even for such very unlikely accidents the plant operators must have the ability and means to mitigate the consequences of such events. This paper outlines a concept for the management of severe accidents based on 1) Computer simulations. 2) Various strategies based on core and containment damage states. 3) Calculational Aids. 4) Procedures. 5) Technical basis report. 6) Training. 7) Drills. The major benefit of this concept is the fact that there is no dedicated operating manual for severe accidents; rather the required mitigative strategies and measures are incorporated into existing accident management manuals leading to truly integrated accident management at the plant. At present this concept is going to be implemented in the NPP Geogen. Although this approach is primarily developed for existing PWRs it is also applicable to other LWRs including new NPP designs. Specific features of the plant can be taken into account by an adaptation of the concept. (authors)

  4. 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.

  5. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

    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

  6. Severe accident insights from the Brunswick IPE

    Energy Technology Data Exchange (ETDEWEB)

    Miller, G.L. (Carolina Power and Light Company, Raleigh, NC (United States))

    1993-01-01

    Insights gained from the development of the level-2 analysis for a Brunswick individual plant examination (IPE) have led to severe accident insights that take advantage of the unique design of the containment structure. The Brunswick steam electric plant (BSEP) consists of two General Electric BWR-4 boiling water reactors (BWRS) with Mark I containments. The containments are unique among BWR Mark I's because the construction of the drywell and torus is reinforced concrete with steel liners. The typical Mark I is a steel shell construction. Both units are rated at 2436 MW(thermal) and [approximately]760 MW(electric). The Brunswick IPE, representing both units, was submitted to the US Nuclear Regulatory Commission in August 1992 (Ref. 1). The estimated mean core damage frequency (CDF) for the level-1 IPE is 2.7 x 10[sup [minus]5]/yr. Station blackout accident sequences contribute 66% to the overall CDF. Transient initiated sequences that involve loss of decay heat removal contribute 30% to the overall CDF. Accident sequences involving anticipated transients without scram (3%), transients with loss of high-pressure injection (I%), loss-of-coolant accidents (LOCAs) (< 1 %), and interfacing LOCAs (< 1 %) constituted the remainder of the accident sequences, which were above the analytical truncation level of 1 X 10 [sup [minus]8]/yr.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  9. 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.

  10. Occupational Radiation Protection in Severe Accident Management

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  12. Nuclear power plant Severe Accident Research Plan

    International Nuclear Information System (INIS)

    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

  13. 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.

  14. Current severe accident research facilities and projects

    International Nuclear Information System (INIS)

    The Working Group on the Analysis and Management of Accidents (GAMA) is mainly composed of technical specialists in the areas of coolant system thermal-hydraulics, in-vessel protection, containment protection, and fission product retention. Its general functions include the exchange of information on national and international activities in these areas, the exchange of detailed technical information, and the discussion of progress achieved in respect of specific technical issues. Severe accident management is one of the important tasks of the group. This document is an update of the 'Current Severe Accident Research Facilities and Projects' list. Facilities and projects are sorted according to the following criteria: In-Vessel Phenomena: Core Degradation and Melt Progression, Molten Core Debris Interaction with the Reactor Pressure Vessel Lower Head and Mechanical Behaviour of Reactor Pressure Vessel Lower Head; In-Vessel and Ex-Vessel Molten Fuel/Coolant Interactions; Ex-Vessel Phenomena: Molten Core Debris/Concrete Interactions, Molten Core/Ceramic Interaction, Melt Release (including DCH), Melt Spreading and Catching Devices Studies, Melt Coolability, Corium Melt properties; Hydrogen Transport and Combustion: Mixing and Distribution, Deflagration, Deflagration-to-Detonation Transition, Passive Recombiner Performance; Mechanical Behaviour of Reactor Pressure Vessel Lower Head; Containment Structural Integrity: Containment Failure Experiment and Analysis, Material Properties and Structural Behaviour, Containment Thermal-Hydraulics, Containment Cooling, Cable Penetration Integrity; Fission Products and Aerosols: Effects of Specific Elements on Iodine Volatility, Release of Low-Volatility Fission Products/Late In-Vessel Fission Product Release, Reactor Materials Release, Aerosol and Iodine Behaviour in Reactor Coolant System and Containment, Retention, Resuspension and Revaporization in Primary Circuit, Aerosol Nucleation and Transport, Source Term, Containment

  15. Ranking of severe accident research priorities

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  17. Sarnet lecture notes on nuclear reactor severe accident phenomenology

    International Nuclear Information System (INIS)

    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)

  18. 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.

  19. 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

  20. 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

  1. Severe Accidents in the Energy Sector

    International Nuclear Information System (INIS)

    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

  2. Porosity effects during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  3. Code strategy for simulating Severe Accident Scenario

    International Nuclear Information System (INIS)

    Severe accident scenarios of Sodium-cooled fast reactors involves various phenomena: core degradation, melt progression towards the core catcher, corium behaviour on the core catcher, energetic corium/sodium interactions, structure mechanical behaviour during expansion phase, containment behaviour, and fission production release and transport. In order to simulate the complete accident scenarios, CEA strategy relies on two sets of calculation codes: a reference set of codes and a set of simplified coupled models dedicated to Probabilistic Risk Assessment analyses. Concerning the reference set, that includes SAS-SFR, SIMMER, CONTAIN, EUROPLEXUS, and TOLBIAC, CEA started, with JAEA and KIT, a validation process based on existing experimental results such as CABRI and SCARABEE programs, and recently against the EAGLE1&2 program results, in the frame of a specific contract with JAEA. Furthermore, CEA is preparing additional experimental programs including in-pile experiments in IGR (NNC reactor), and out-of-pile experiments in the future experimental FOURNAISE facility to be built in CEA Cadarache (France). (author)

  4. SARNET: Severe accident research network of excellence

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  5. 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.

  6. Aerosol transport in severe reactor accidents

    International Nuclear Information System (INIS)

    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. Consideration of severe accidents in design of advanced WWER reactors

    International Nuclear Information System (INIS)

    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

  8. Severe accident analysis to verify the effectiveness of severe accident management guidelines for large pressurized heavy water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Gokhale, O.S., E-mail: onkarsg@barc.gov.in; Mukhopadhyay, D., E-mail: dmukho@barc.gov.in; Lele, H.G., E-mail: hglele@barc.gov.in; Singh, R.K., E-mail: rksingh@barc.gov.in

    2014-10-15

    Highlights: • The progression of severe accident initiated from high pressure scenario of station black out has been analyzed using RELAP5/SCDAP. • The effectiveness of SAMG actions prescribed has been established through analysis. • The time margin available to invoke the SAMG action has been specified. - Abstract: The pressurized heavy water reactor (PHWR) contains both inherent and engineered safety features that help the reactor become resistant to severe accident and its consequences. However in case of a low frequency severe accident, despite the safety features, procedural action should be in place to mitigate the accident progression. Severe accident analysis of such low frequency event provides insight into the accident progression and basis to develop the severe accident management guidelines (SAMG). Since the order of uncertainty in the progression path of severe accident is very high, it is necessary to study the consequences of the SAMG actions prescribed. The paper discusses severe accident analysis for large PHWRs for multiple failure transients involving a high pressure scenario (initiation event like SBO with loss of emergency core cooling system and loss of moderator cooling). SAMG actions prescribed for such a scenario include water injection into steam generator, calandria vessel or calandria vault at different stages of accident. The effectiveness of SAMG actions prescribed has been investigated. It is found that there is sufficient time margin available to the operator to execute these SAMG actions and the progression of severe accident is arrested in all the three cases.

  9. Comparative Assessment of Severe Accidents in the Chinese Energy Sector

    International Nuclear Information System (INIS)

    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)

  10. 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)

  11. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    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)

  12. Iodine chemical forms in LWR severe accidents

    International Nuclear Information System (INIS)

    Calculated data from seven severe accident sequences in light water reactor plants were used to assess the chemical forms of iodine in containment. In most of the calculations for the seven sequences, iodine entering containment from the reactor coolant system was almost entirely in the form of CsI with very small contributions of I or HI. The largest fraction of iodine in forms other than CsI was a total of 3.2% as I plus HI. Within the containment, the CsI will deposit onto walls and other surfaces, as well as in water pools, largely in the form of iodide (I-). The radiation-induced conversion of I- in water pools into I2 is strongly dependent on pH. In systems where the pH was controlled above 7, little additional elemental iodine would be produced in the containment atmosphere. When the pH falls below 7, it may be assumed that it is not being controlled and large fractions of iodine as I2 within the containment atmosphere may be produced. 17 refs., 5 tabs

  13. On severe accident hydrogen behaviour in Loviisa

    International Nuclear Information System (INIS)

    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.)

  14. Iodine chemical forms in LWR severe accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-01-01

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

  15. 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.).

  16. 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.

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

    International Nuclear Information System (INIS)

    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

  18. Development of severe accident management and training support system

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

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

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

    OpenAIRE

    Tusheva, Polina

    2013-01-01

    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 re...

  20. Severe Accident Management Strategy for EU-APR1400

    International Nuclear Information System (INIS)

    In EU-APR1400, the dedicated instrumentation and mitigation features for SAM are being developed to keep the integrity of containment and to prevent the uncontrolled release of fission products. In this paper, SAM strategy for EU-APR1400 was introduced in stages. It is still under development and finally the Severe Accident Management Guidance will be completed based on this SAM Strategy. Severe accidents in a nuclear power plant are defined as certain unlikely event sequences involving significant core damage with the potential to lead to significant releases according to EUR 2.1.4.4. Even though the probability of severe accidents is extremely low, the radiation release may cause serious effect on people as well as environment. Severe Accident Management (SAM) encompasses those actions which could be considered in recovering from a severe accident and preventing or mitigating the release of fission products to the environment. Whether those actions are successful or not, depending on a progression status of a severe accident to mitigate the consequences of severe accident phenomena to limit the release of radioactive materials keeping the leak tightness of the Primary Containment, and finally to restore transient severe accident progression into a controlled and safe states

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  4. MELCOR simulation of postulated severe accidents in OPR1000

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-10-15

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

  5. A study on PHWR moderator and severe accident analysis system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyoung Tae; Rhee, B. W.; Kim, D. H. [KAERI, Daejeon (Korea, Republic of); and others

    2012-04-15

    For the purpose of establishment of PHWR moderator and severe accident analysis system, the following works are performed. The main thermal-hydraulic phenomena are investigated and scaling analysis of the scaled down test facility design and fabrication are done to determine the scaling ratio based on the scaling law and practical constraints of the test facility. Theoretical background of the commercial CFD codes has been found out and their applicability and application conditions for the moderator circulation analysis are reviewed to develop the computer code requirement for the moderator 3-D analysis codes. Satisfactory analysis results against the STERN Lab. experiment showed the applicability of OpenFOAM and CUPID codes to moderator circulation analysis. For the development of various accident scenarios for establishing the DB for severe accident phenomena/progression, the level 1 and the level 2 PSA analysis results for Wolsong Unit 1 are reviewed and the most probable accident scenarios from the PDS event trees are selected. The latest ISAAC 4.03 version is used to predict the basic accident progression and the improvement items for the most up-to-date severe accident analysis issues analyzing function are derived. A basic system for the PHWR severe accident management decision making support system, SAMEX-CR is set up and requirement for the DB management system, SARDB-CR is derived to develop the implementation methodology for severe accident analysis DB management system.

  6. The philosophy of severe accident management in the US

    International Nuclear Information System (INIS)

    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)

  7. A knowledge based severe accident handbook for PWR

    International Nuclear Information System (INIS)

    During the last decade the level of knowledge about severe accident phenomena has increased dramatically. The improved understanding has been achieved by extensive research but also from feed-back of experience from actual incidents/accidents such as Three Mile Island and Chernobyl. In Sweden, mitigating measures such as filtered venting and external water source were implemented at all nuclear power plants by 1988. In parallel the Emergency Operating Procedures (at Ringhals called Emergency Response Guidelines, ERG, and Beyond ERG, BERG) were developed to include these new features. However, the accident management system has since then been further improved and one important aspect is the long-term accident management. The new information obtained has been one of the basis for a new knowledge based handbook to support the unit leader and the Technical Support Center. The handbook contains information concerning specific issues in the BERG and advice how the organization can manage a long-term severe accident situation

  8. 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).

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

    International Nuclear Information System (INIS)

    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

  10. MELCOR DB Construction for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  11. Severities of transportation accidents involving large packages

    International Nuclear Information System (INIS)

    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

  12. Severities of transportation accidents involving large packages

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

    The ISAAC ((Integrated Severe Accident Analysis Code for CANDU Plant) code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a Level 2 probabilistic safety assessment or severe accident management strategy developments. The code has the capability to predict a severe accident progression by modeling the CANDU6- specific systems and the expected physical phenomena based on the current understanding of the unique accident progressions. The code models the sequence of accident progressions from a core heatup, pressure tube/calandria tube rupture after an uncovery from inside and outside, a relocation of the damaged fuel to the bottom of the calandria, debris behavior in the calandria, corium quenching after a debris relocation from the calandria to the calandria vault and an erosion of the calandria vault concrete floor, a hydrogen burn, and a reactor building failure. Along with the thermal hydraulics, the fission product behavior is also considered in the primary system as well as in the reactor building.

  17. Drug use and the severity of a traffic accident

    NARCIS (Netherlands)

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

    2005-01-01

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

  18. Severe accidents in nuclear power plants. V.2

    International Nuclear Information System (INIS)

    The International Symposium on Severe Accidents in Nuclear Power Plants, organized by the International Atomic Energy Agency and co-sponsored by the Nuclear Energy Agency of the OECD, was held in Sorrento, Italy, from 21 to 25 March 1988. The symposium was attended by over 300 participants from 35 Member States and 4 organizations. There were 72 oral presentations and 28 poster presentations. In addition, a special session devoted to the publication entitled Basic Safety Principles for Nuclear Power Plants was organized by the International Nuclear Safety Advisory Group (INSAG) in the form of a panel discussion. The objective of the symposium was to provide a forum for an international exchange of information on the scientific and technical aspects of severe accidents, and on the rationale and implementation of severe accident practices in participating countries. The papers provided an excellent overview of different national approaches, with the overall emphasis on preventive, mitigative and accident management measures. Every reasonable effort is being made in design and operation to prevent accidents from happening and to limit the consequences of any that might occur. However, it is also generally considered prudent to introduce design modifications and operational changes and prepare contingency plans for dealing with a possible accident. The actual measures taken vary from country to country but usually involve detailed extended or new emergency operating procedures and the use of existing and/or new systems to limit off-site releases. Containment filtering and venting, the use of mobile equipment and the utilization of external water sources were among the options presented and discussed in detail. This is volume 2 of the proceedings of a symposium. Two main scientific and technical topics are presented in this volume: accident research and development (34 papers) and accident management (24 papers). A separate abstract was prepared for each of these papers

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

    International Nuclear Information System (INIS)

    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

  20. 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)

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

    International Nuclear Information System (INIS)

    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

  2. Severe accidents due to windsurfing in the Aegean Sea.

    Science.gov (United States)

    Kalogeromitros, A; Tsangaris, H; Bilalis, D; Karabinis, A

    2002-06-01

    Windsurfing is a popular sport and has recently become an Olympic event. As an open-air water activity that requires the participant to be in perfect physical condition, windsurfers may be prone to accidents when certain basic rules or procedures are violated. The current study monitored severe injuries due to windsurfing over a period of 12 months in the Aegean Sea in Greece. Our study revealed 22 cases of severe accidents due to windsurfing, with a wide range of injuries including head injuries, spinal cord injuries, and severe fractures of the extremities. Prolonged hospitalization, severe disability and two deaths occurred as consequences of these accidents. The study examined the characteristics of these patients and the possible risk factors and conditions associated with the accidents. We also focused on the most common types of injuries and reviewed the mechanisms that may provoke them. Water sports and particularly windsurfing represent a major challenge for the emergency medical system, especially in the Aegean Sea. Hundreds of islands, kilometres of isolated coasts, millions of tourists, an extended summer period and rapidly changing weather create conditions that constantly test the efficacy of the emergency services. The development of an appropriate infrastructure and maximum control of the risk factors causing these accidents could reduce the morbidity and mortality that, unfortunately but rather predictably, accompany this popular summer activity. PMID:12131638

  3. A framework for assessing severe accident management strategies

    International Nuclear Information System (INIS)

    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

  4. Severe accidents in nuclear power plants. V.1

    International Nuclear Information System (INIS)

    The International Symposium on Severe Accidents in Nuclear Power Plants, organized by the International Atomic Energy Agency and co-sponsored by the Nuclear Energy Agency of the OECD, was held in Sorrento, Italy, from 21 to 25 March 1988. The symposium was attended by over 300 participants from 35 Member States and 4 organizations. There were 72 oral presentations and 28 poster presentations. In addition, a special session devoted to the publication entitled Basic Safety Principles for Nuclear Power Plants was organized by the International Nuclear Safety Advisory Group (INSAG) in the form of a panel discussion. The objective of the symposium was to provide a forum for an international exchange of information on the scientific and technical aspects of severe accidents, and on the rationale and implementation of severe accident practices in participating countries. All the presentations were divided into three chapters: National positions and practices on severe accidents (14 papers); Accident initiation and analysis (21 papers); Non-water cooled power reactors (5 papers). A separate abstract was prepared for each of these papers. Refs, figs and tabs

  5. 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

  6. Challenges in thermohydraulic analysis of LWR severe accidents: steam explosions

    International Nuclear Information System (INIS)

    A severe accident is an accident state beyond design basis events with significant core damage and release of radioactive materials to the environment. Nuclear power plants are designed to endure prescribed accident situations against which safety equipment should be effective enough to assure that environmental release of radioactive materials is avoided. However, three major severe accidents have already experienced in commercial scale power plants so far, namely, Three Mile Island (TMI), Chernobyl and Fukushima Daiichi. Thus, the severe accident is no more just a hypothesis but a reality that have to be prepared with enough effectivity. A method for assessment of steam explosion load has been established based on presently available phenomenological information and simulation technique. On the 3 other hand, the present model is not sufficient for slow long term FCIs in which the steam and non-condensable gas generation rate for vessel pressurization and the resulting debris bed geometry for its coolability are in question. Also, there are shortcomings from the present analytical method such as influences of the mesh size on the void fraction, lacking radiation heat transfer beyond meshes and so on. If the level of the model is upgraded to CFD type including more flexible particle methods, direct simulation of complicated phenomena involving molten core, may become available. This may be one of the directions of future development

  7. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    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

  8. 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

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

    International Nuclear Information System (INIS)

    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

  10. 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

  11. 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.

  12. Severe accident analysis code Sampson for impact project

    Energy Technology Data Exchange (ETDEWEB)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh [Nuclear Power Engineering Corporation, Advanced Simulation Systems Dept., Tokyo (Japan)

    2001-07-01

    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)

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

    OpenAIRE

    Martin, Robert P.

    2012-01-01

    A general evaluation methodology development and application process (EMDAP) paradigm is described for the resolution of severe accident safety issues. For the broader objective of complete and comprehensive design validation, severe accident safety issues are resolved by demonstrating comprehensive severe-accident-related engineering through applicable testing programs, process studies demonstrating certain deterministic elements, probabilistic risk assessment, and severe accident management...

  14. 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.

  15. ANS severe accident program overview ampersand planning document

    International Nuclear Information System (INIS)

    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

  16. Aerosol Characterization in Containment Air during Severe Accident

    International Nuclear Information System (INIS)

    To ensure the reduction of the radioactive aerosol concentration and to guarantee the filter efficiency in accident scenarios with various conditions, it is essential to characterize the aerosols in the containment air. This study is to investigate the aerosol size distribution and the concentration in containment air during the severe accident scenario by using numerical simulations. NAUA code was used to model the behavior of radioactive aerosol particles. As input parameters for NAUA simulation, the data of the currently operating nuclear power plant (OPR-1000) was used and conservative thermal hydraulic conditions were provided from the conservative simulation results. For verification, the simulation results were compared with the data found in the literature. Aerosol in containment air during severe accident is modeled by using NAUA code. The aerosol characteristics are calculated and variations due to some parameters are investigated. For verification, the main results are compared with the information of the previous works. The simulation results in this study for particle size distribution in containment air during severe accident were in general agreement with previously reported measurements. The simulation results and findings would be useful data for prototypic CFVS design and for planning further experimental studies

  17. Severe accident natural circulation studies at the INEL

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

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

    International Nuclear Information System (INIS)

    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 CO2-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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

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

    International Nuclear Information System (INIS)

    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

  1. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, L.E.; Chou, C.K.; Gerhard, M.A.; Kimura, C.Y.; Martin, R.W.; Mensing, R.W.; Mount, M.E.; Witte, M.C.

    1987-02-01

    Volume 2 contains the following appendices: Severe accident data; truck accident data; railroad accident data; highway survey data and bridge column properties; structural analysis; thermal analysis; probability estimation techniques; and benchmarking for computer codes used in impact analysis. (LN)

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

    International Nuclear Information System (INIS)

    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. Proceedings of the specialist meeting on severe accident management implementation

    International Nuclear Information System (INIS)

    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

  5. Hydrogen distribution during postulated severer accident in kaiga containment

    Energy Technology Data Exchange (ETDEWEB)

    Sanjeev Kumar; Manoj Kansal; Nalini Mohan; Bhawal, R.N.; Bajaj, S.S. [Nuclear Power Corporation of India Limited ENT-1, R-3, Nub Anushakti Nagar, Mumbai-400094 (India)

    2005-07-01

    Full text of publication follows: Generation and accumulation of hydrogen in containment atmosphere during postulated accident scenario could pose a potential threat to the integrity of the containment as the hydrogen can form flammable or even explosive mixture with air in the containment. The governing accident sequence considered for this evaluation is a dual failure involving a double-ended break in reactor inlet header in Fuelling Machine Vault (FMV) with unavailability of Emergency Core Cooling System (ECCS). Consequences of any break of Primary Heat Transport (PHT) boundary in pump room would be less severe compared with that in FMV because the hydrogen releasing during such accident scenario would be directly mixing with the volume of pump room, which is several times (15 times) higher than FMV and hence may result low local hydrogen concentration in comparison to FMV. In case of reactor header break, the hydrogen generated due to metal water reaction is expected to be released to Fuelling Machine Vault(Break Compartment) and mix uniformly with air and steam in the vault. Subsequently, additional hydrogen is expected to be released to suppression pool at a slower rate due to radiolysis of pool water. As total of amount hydrogen generation is not much, the global concentration of hydrogen would not reach at flammability limit of 4% even after 10 days of accident. The local concentration in break compartment (FMV) may cross the flammability limit or even detonation limit during initial period of accident as hydrogen generation rate would be very high due to metal water reaction. To study the hydrogen distribution and to limit the local hydrogen concentration in various compartments of containment during postulated accident, the analyses were carried out by providing the forced circulation between pump room and FM vaults. Analyses were also repeated by stopping reactor-building coolers in some selected areas. The study reveals that under postulated severe

  6. Hydrogen distribution during postulated severer accident in kaiga containment

    International Nuclear Information System (INIS)

    Full text of publication follows: Generation and accumulation of hydrogen in containment atmosphere during postulated accident scenario could pose a potential threat to the integrity of the containment as the hydrogen can form flammable or even explosive mixture with air in the containment. The governing accident sequence considered for this evaluation is a dual failure involving a double-ended break in reactor inlet header in Fuelling Machine Vault (FMV) with unavailability of Emergency Core Cooling System (ECCS). Consequences of any break of Primary Heat Transport (PHT) boundary in pump room would be less severe compared with that in FMV because the hydrogen releasing during such accident scenario would be directly mixing with the volume of pump room, which is several times (15 times) higher than FMV and hence may result low local hydrogen concentration in comparison to FMV. In case of reactor header break, the hydrogen generated due to metal water reaction is expected to be released to Fuelling Machine Vault(Break Compartment) and mix uniformly with air and steam in the vault. Subsequently, additional hydrogen is expected to be released to suppression pool at a slower rate due to radiolysis of pool water. As total of amount hydrogen generation is not much, the global concentration of hydrogen would not reach at flammability limit of 4% even after 10 days of accident. The local concentration in break compartment (FMV) may cross the flammability limit or even detonation limit during initial period of accident as hydrogen generation rate would be very high due to metal water reaction. To study the hydrogen distribution and to limit the local hydrogen concentration in various compartments of containment during postulated accident, the analyses were carried out by providing the forced circulation between pump room and FM vaults. Analyses were also repeated by stopping reactor-building coolers in some selected areas. The study reveals that under postulated severe

  7. 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.

  8. Safety against releases in severe accidents. Final report

    International Nuclear Information System (INIS)

    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. Investigation of Key Factors for Accident Severity at Railroad Grade Crossings by Using a Logit Model

    OpenAIRE

    Hu, Shou-Ren; Li, Chin-Shang; Lee, Chi-Kang

    2010-01-01

    Although several studies have used logit or probit models and their variants to fit data of accident severity on roadway segments, few have investigated accident severity at a railroad grade crossing (RGC). Compared to accident risk analysis in terms of accident frequency and severity of a highway system, investigation of the factors contributing to traffic accidents at an RGC may be more complicated because of additional highway–railway interactions. Because the proportional odds assumption ...

  10. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    Energy Technology Data Exchange (ETDEWEB)

    R. A. Wigeland; J. E. Cahalan

    2009-12-01

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

  11. Westinghouse severe accident management guidance overview and current status

    International Nuclear Information System (INIS)

    The Westinghouse Owners Group has completed a major development program in Severe Accident Management. This program draws on all presently available sources of information in the field, including in the field, including NRC, NUMARC and EPRI programs, plant specific Individual Plant Examinations and Probabilistic Safety Assessments, and other international activities. The program has developed a full set of Severe Accident Management Guidance (SAMG) applicable to Westinghouse and Westinghouse licensee PWR plant. The SAMG enhances the capabilities of the plant emergency response team for accident sequences that progress to fuel damage, and therefore beyond the range of applicability of present guidance in the form of Emergency Operating Procedures. Since the first draft of SAMG was transmitted officially to the WOG members and the NRC in July 1993, many activities have been carried out by the different organizations involved, and although no significant changes to the SAMG structure have resulted from these activities, several enhancement have been included, mainly from the comments recorded during the generic SAMG validation exercise at the Point Beach plant. With the issue in June 1994 of the revision 0 SAMG, some plants in the U.S. and abroad are already implementing plant specific guidelines. This paper provides an overview of the SAMG package, and also describe the most important comments and feedback from the validation and review efforts. (author)

  12. Radionuclide release calculations for selected severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Denning, R.S.; Leonard, M.T.; Cybulskis, P.; Lee, K.W.; Kelly, R.F.; Jordan, H.; Schumacher, P.M.; Curtis, L.A. (Battelle Columbus Div., OH (USA))

    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.

  13. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

    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

  14. Steam Oxidation Testing in the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-01

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

  15. Essential severe accident mitigation measures for operating and future PWR's

    International Nuclear Information System (INIS)

    Severe Accident mitigation measures are a constituent of the safety concept in Europe not only for operating but also for future light water reactors. While operating reactors mainly have been backfitted with such measure, for future reactors Severe Accident mitigation measures already have to be considered in the design phase. Severe Accident measures are considered as the 4th level of defense for future reactors. This difference has consequences also on the kind of measures proposed to be introduced. While in operating plants Severe Accident mitigation measures are considered for further risk reduction, in future reactors an explicit higher level of safety is required resulting in additional design measures. This higher safety level is expressed in the requirement that there must be no need for evacuation of surrounding populations except in the immediate vicinity of the plant and for long-term restrictions with regard to the consumption of locally grown food. Because of the potential hazard posed by radioactive releases to the environment in the event of an Severe Accident situation depends largely on the airborne material in the containment atmosphere and on the containment integrity, new system features to prevent loss of containment integrity have been introduced in the design of the NPP's. For these tasks it has been necessary to develop and qualify new system technologies and implement them finally into NPP's, e.g. like systems for containment atmosphere H2-control, filtered venting, core retention devices and atmosphere sampling. The following systems are introduced for operating as well as for future plants: · The Hydrogen Control System is based on the Passive Autocatalytic Recombiner (PAR) technology. There is no need for any operator actions because of the self-starting feature of the catalyst if hydrogen is released. · In situ Post Accident Sampling System (In situ-PASS) are introduced for the purpose of obtaining realistic information on airborne

  16. Timing of core damage states following severe accidents for the CANDU reactor design

    International Nuclear Information System (INIS)

    This paper documents the analytical methodology used to evaluate severe accident sequences. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression are described. An example of a specific accident scenario is provided in order to illustrate the application of the severe accident progression methodology. The postulated sever accidents analyzed mainly differ in the timing to reach and progress through each defined 'core damage state'. (author)

  17. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

    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)

  18. 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.

  19. 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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

  2. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    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)

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

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    light. Occurrence of injury to bus passengers is positively related to (i) the involvement of heavy vehicles, (ii) crossing intersections in yellow or red light, (iii) open areas, (iv) high speed limits, and (v) slippery road surface. Conclusions The findings of the current study provide a comprehensive...... 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...

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

    International Nuclear Information System (INIS)

    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

  5. 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

  6. Severe head injury caused by motorcycle traffic accident

    Institute of Scientific and Technical Information of China (English)

    李钢

    1999-01-01

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

  7. Development Process of Plant-specific Severe Accident Management Guidelines for Wolsong Nuclear Power Plants

    International Nuclear Information System (INIS)

    A severe accident, which occurred at the TMI in 1979 and Chernobyl in 1986, is an accident that exceeds design basis accidents and leads to significant core damage. The severe accident is the low possibility of occurrence but the high severity. To mitigate the consequences of the severe accidents, Korean Nuclear Safety Committee declared the Severe Accident Policy in 2001, which requested the development of Severe Accident Management Guidelines (SAMGs) for operating plants. SAMG is a symptom-based guidance that takes a set of actions to alleviate the outcomes of severe accidents and to get into the safe stable plant condition. The purpose of this paper is to presents the strategic development process of the PHWR SAMG. The guidelines consist of 5 categories: an emergency guide for the main control room (MCR) operators, a strategy implementing guide for the technical support center (TSC), six mitigation guides, a monitoring guide, and a termination guide

  8. 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

  9. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

    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

  10. 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)

  11. Advances in operational safety and severe accident research

    International Nuclear Information System (INIS)

    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)

  12. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

  14. Root Causes and Impacts of Severe Accidents at Large Nuclear Power Plants

    OpenAIRE

    Högberg, Lars

    2013-01-01

    The root causes and impacts of three severe accidents at large civilian nuclear power plants are reviewed: the Three Mile Island accident in 1979, the Chernobyl accident in 1986, and the Fukushima Daiichi accident in 2011. Impacts include health effects, evacuation of contaminated areas as well as cost estimates and impacts on energy policies and nuclear safety work in various countries. It is concluded that essential objectives for reactor safety work must be: (1) to prevent accidents from d...

  15. Shipping container response to three severe railway accident scenarios

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  18. Application of probabilistic safety assessment in CPR1000 severe accident prevention and mitigation analysis

    International Nuclear Information System (INIS)

    The relationship between probabilistic safety assessment (PSA) and severe accident study was discussed. Also how to apply PSA in severe accident prevention and mitigation was elaborated. PSA can find the plant vulnerabilities of severe accidents prevention and mitigation. Some modifications or improvements focusing on these vulnerabilities can be put forward. PSA also can assess the efficient of these actions for decision-making. According to CPR1000 unit severe accident analysis, an example for the process and method on how to use PSA to enhance the ability to deal with severe accident prevention and mitigation was set forth. (authors)

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

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-15

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

  1. Role of accident analysis in development of severe accident management guidance for multi-unit CANDU nuclear power plants

    International Nuclear Information System (INIS)

    This paper discusses the role of accident analysis in support of the development of Severe Accident Management Guidance for domestic CANDU reactors. In general, analysis can identify what types of challenges can be expected during accident progression but it cannot specify when and to what degree accident phenomena will occur. SAMG overcomes these limitations by monitoring the actual values of key plant indicators that can be used directly or indirectly to infer the condition of the plant and by establishing setpoints beyond which corrective action is required. Analysis can provide a means to correlate observed post-accident plant behavior against predicted behaviour to improve the confidence in and quality of accident mitigation decisions. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-07-01

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

  3. A synthesis of hydrogen behaviour in severe reactor accidents

    International Nuclear Information System (INIS)

    The report discusses hydrogen behaviour in severe reactor accidents. An attempt is made to investigate most of the hydrogen-related phenomena into a synthetic approach. Simplistic analytical methods are developed and applied to the Finnish nuclear power plants, yielding easily understandable and reproducible results. The core dryout, heatup and melting phases are first described, focusing on hydrogen production during core degradation and fuel-coolant interactions. Hydrogen production is quantified, and the most important determinants are identified. The potential pressure and temperature loads caused by hydrogen in the containment calculated, and the need for mitigative measures is assessed. Finally, the natural circulation and mixing of containment gases are touched upon in order to estimate the potential success of proposed hydrogen mitigation measures. (39 refs., 6 fig., 27 tabs.)

  4. Present status of research activities in severe accident evaluation for nuclear power plants

    International Nuclear Information System (INIS)

    The basis for securing nuclear safety is to prevent occurrence of accidents and to mitigate propagation of abnormal events or accidents to severe accidents. In practice, a nuclear power plant is designed and constructed so that abnormal events can be detected at the early phase to cope with the events and safety features and facilities are installed to mitigate and reduce the consequences in the case of such accidents. However it is important to prepare preventive measures as well as mitigative measures to cope with severe accidents to further improve the level of safety. Research on the evaluation of severe accidents is needed to develop such measures. Severe accident research is performed in many countries including Japan and a lot of findings have been made. At JAERI, experiments are being conducted to clarify severe accident phenomena and to make quantitative evaluation of safety margin of a nuclear power plant against severe accidents. A lot of findings on the fuel damage process in the early phase of severe accidents have been obtained in the past years. However there are still large uncertainties on the fuel damage process in the late phase of accidents. In the area of accident management, there exists need for experiments and analyses. (author)

  5. Analysis of severe accidents in the IIE - Instituto de Investigaciones Electricas

    International Nuclear Information System (INIS)

    The international trend on several accident analysis shows an overall emphasis on prevention, mitigation and management of severe accidents in nuclear power plants. Most of the developed countries have established policies and programs to deal with accidents beyond design basis. An encouraged participation in severe accidents analysis of the Latin American Countries operating commercial Nuclear Power Plants is forseen. The experience from probabilistic safety assessment, emergency operating procedures and best estimate codes for transient analysis, in order to develop analysis tools and knowledge that support the severe accident programs of the national nuclear power organizations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-15

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

  7. Study on severe accident mitigation measures for the development of PWR SAMG

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

    In the development of the Severe Accident Management Guidelines (SAMG), it is very important to choose the main severe accident sequences and verify their mitigation measures. In this article, Loss-of-Coolant Accident (LOCA), Steam Generator Tube Rupture (SGTR), Station Blackout (SBO), and Anticipated Transients without Scram (ATWS) in PWR with 300 MWe are selected as the main severe accident sequences. The core damage progressions induced by the above-mentioned sequences are analyzed using SCDAP/RELAP5. To arrest the core damage progression and mitigate the consequences of severe accidents, the measures for the severe accident management (SAM) such as feed and bleed, and depressurizations are verified using the calculation. The results suggest that implementing feed and bleed and depressurization could be an effective way to arrest the severe accident sequences in PWR.

  8. Evaluation of severe accident risks: Quantification of major input parameters: MAACS [MELCOR Accident Consequence Code System] input

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

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

    International Nuclear Information System (INIS)

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

  15. Proceedings of the specialist meeting on selected containment severe accident management strategies

    International Nuclear Information System (INIS)

    Twenty papers were presented at the first specialist meeting on Selected Containment Severe Accident management Strategies, held in Stockholm, Sweden, in 1994, half of them dealing with accident management strategies implementation status, half of them with research aspects. The four sessions were: general aspects of containment accident management strategies, hydrogen management techniques, other containment accident management strategies (spray cooling, core catcher...), surveillance and protection of containment function

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    management tool.Initially models were built by using existing traffic accident data collected by the police and emergency rooms in Denmark. The data registered by the police was collected on traffic accidents occurred on Danish roads in the period between 2002 and 2008. The emergency room data were collected...... to be the most relevant factor related to the lack of intention to report future cycling accidents. Secondly, the factors: concerns about family distress and social image and preference to allocate time to other activities are both associated with non-reporting intentions (Paper 3). 5) New information about...

  19. Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

    OpenAIRE

    Min Yoo; Sung Min Shin; Hyun Gook Kang

    2014-01-01

    Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently pose...

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

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

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

  6. Severe accident risks: An assessment for five US nuclear power plants

    International Nuclear Information System (INIS)

    This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United State. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide releases and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. This report, Volume 3, contains two appendices. Appendix D summarizes comments received, and staff responses, on the first (February 1987) draft of NUREG-1150. Appendix E provides a similar summary of comments and responses, but for the second (June 1989) version of the report

  7. Severe accident risks: An assessment for five US nuclear power plants: Appendices A, B, and C

    International Nuclear Information System (INIS)

    This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United States. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two or the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide release and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. Volume 2 of this report contains three appendices, providing greater detail on the methods used, an example risk calculation, and more detailed discussion of particular technical issues found important in the risk studies

  8. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

    Iwashita, Tsuyoshi; Ujita, Hiroshi [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan)

    2000-11-01

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

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

    OpenAIRE

    Jan Christian Kaiser

    2012-01-01

    Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES) level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI) 4; 62) severe accidents am...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  12. Review of current Severe Accident Management (SAM) approaches for Nuclear Power Plants in Europe

    OpenAIRE

    HERMSMEYER Stephan; Iglesias, R.; Herranz, L; REER B.; SONNENKALB M; NOWACK H.; Stefanova, A.; Raimond, E.; CHATELARD P.; FOUCHER Laurent; BARNAK M.; MATEJOVIC P; PASCAL GHISLAIN; VELA GARCIA MONICA; SANGIORGI MARCO

    2014-01-01

    The Fukushima accidents highlighted that both the in-depth understanding of such sequences and the development or improvement of adequate Severe Accident Management (SAM) measures are essential in order to further increase the safety of the nuclear power plants operated in Europe. To support this effort, the CESAM (Code for European Severe Accident Management) R&D project, coordinated by GRS, started in April 2013 for 4 years in the 7th EC Framework Programme of research and development of th...

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

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1998-05-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-12-15

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Gunhyo; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2015-03-15

    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

  2. French regulatory requirements concerning severe accidents in PWRs and associated research programme

    International Nuclear Information System (INIS)

    This report gives a global view of the French reactor safety approach; aspects in relation with severe accidents are pointed out: safety goals regarding population, and safety goals regarding plant design. Ultimate or U procedures involving physical phenomena of severe accidents are then described. R. and D. programs have been defined with regard to the priorities resulting from this approach

  3. Mitigation of Hydrogen Hazards in Severe Accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

    Consideration of severe accidents in nuclear power plants is an essential component of the defence in depth approach in nuclear safety. Severe accidents have very low probabilities of occurring, but may have significant consequences resulting from the degradation of nuclear fuel. The generation of hydrogen and the risk of hydrogen combustion, as well as other phenomena leading to overpressurization of the reactor containment in case of severe accidents, represent complex safety issues in relation to accident management. The combustion of hydrogen, produced primarily as a result of heated zirconium metal reacting with steam, can create short term overpressure or detonation forces that may exceed the strength of the containment structure. An understanding of these phenomena is crucial for planning and implementing effective accident management measures. Analysis of all the issues relating to hydrogen risk is an important step for any measure that is aimed at the prevention or mitigation of hydrogen combustion in reactor containments. The main objective of this publication is to contribute to the implementation of IAEA Safety Standards, in particular, two IAEA Safety Requirements: Safety of Nuclear Power Plants: Design and Safety of Nuclear Power Plants: Operation. These Requirements publications discuss computational analysis of severe accidents and accident management programmes in nuclear power plants. Specifically with regard to the risk posed by hydrogen in nuclear power reactors, computational analysis of severe accidents considers hydrogen sources, hydrogen distribution, hydrogen combustion and control and mitigation measures for hydrogen, while accident management programmes are aimed at mitigating hydrogen hazards in reactor containments.

  4. Knowledge data base for severe accident management of nuclear power plants

    International Nuclear Information System (INIS)

    For the reinforcement of the safety of NPPs, the continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of this present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of severe accident, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of accident management. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the severe accident analysis codes and the accident management knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2010 are as follows; Experimental study on OECD/NEA projects such as MCCI, SERENA, SFP and international cooperative PSI-ARTIST project, and analytical study on accident management review of new plant and making regulation for severe accident. (author)

  5. Knowledge data base for severe accident management of nuclear power plants

    International Nuclear Information System (INIS)

    For the reinforcement of the safety of NPPs, the continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of this present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of severe accident, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of accident management. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the severe accident analysis codes and the accident management knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2011 are as follows; Experimental study on OECD/NEA projects such as MCCI, SERENA, SFP and international cooperative PSI-ARTIST project, and analytical study on accident management review of new plant and making regulation for severe accident. (author)

  6. Specialist meeting on selected containment severe accident management strategies. Summary and conclusions

    International Nuclear Information System (INIS)

    The CSNI Specialist Meeting on Selected Containment Severe Accident Management Strategies held in Stockholm, Sweden in June 1994 was organised by the Task Group on Containment Aspects of Severe Accident Management (CAM) of CSNI's Principal Working Group on the Confinement of Accidental Radioactive Releases (PWG4) in collaboration with the Swedish Nuclear Power Inspectorate (SKI). Conclusions and recommendations are given for each of the sessions of the workshops: Containment accident management strategies (general aspects); hydrogen management techniques and other containment accident management techniques; surveillance and protection of containment function

  7. Comparison of selected U.S. highway and railway severe accidents to U.S. regulatory accident conditions and IAEA transport standards

    International Nuclear Information System (INIS)

    This paper discusses selected severe historical US highway and rail accidents and compares the mechanical and/or thermal environments associated with these accidents to the 10CFR71 Hypothetical Accident Conditions and the accident environments (both regulatory and extraregulatory) investigated in 'Shipping Container Response to Severe Highway and Railway Accident Conditions', which is commonly known as the Modal Study, and in 'Re-examination of Spent Fuel Shipment Risk Estimates', NUREG/CR-6672. Since the hypothetical accident conditions of 10CFR71 are similar to the International Atomic Energy Agency's (IAEA) package tests for accident conditions of transport, the evaluation is also valid in demonstrating the adequacy of IAEA's transport safety standard. Careful examination of the reports on the severe accidents revealed the accidents were found to be bounded by the regulatory environment described in 10CFR71. (author)

  8. Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

    International Nuclear Information System (INIS)

    Effective Accident Management planning can produce both a reduction in the frequency of severe accidents at nuclear power plants as well as the ability to mitigate a severe accident. The purpose of an accident management programme is to provide to the responsible plant staff the capability to cope with the complete range of credible severe accidents. This requires that appropriate instrumentation and equipment are available within the plant to enable plant staff to diagnose the faults and to implement appropriate strategies. The programme must also provide the necessary guidance, procedures, and training to assure that appropriate corrective actions will be implemented. One of the key issues to be discussed is the transition from control room operations and the associated emergency operating procedures to a technical support team approach (and the associated severe accident management strategies). Following a proposal made by the Senior Group of Experts on Severe Accident Management (SESAM), the Committee on the Safety of Nuclear Installations decided to sponsor a Specialist Meeting on Severe Accident Management Programme Development. The general objectives of the Specialist Meeting were to exchange experience, views, and information among the participants and to discuss the status of severe accident management programmes. The meeting brought together utilities, accident management programme developers, personnel training programme developers, regulators, and researchers. In general, the tone of the Specialist Meeting - designed to promote progress, as contrasted with conferences or symposia where the state-of-the-art is presented - was to be rather practical, and focus on accident management programme development, applications, results, difficulties and improvements. As shown by the conclusions of the meeting, there is no doubt that this objective was widely attained

  9. 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)

  10. Shipping container response to severe highway and railway accident conditions: Main report

    International Nuclear Information System (INIS)

    This report describes a study performed by the Lawrence Livermore National Laboratory to evaluate the level of safety provided under severe accident conditions during the shipment of spent fuel from nuclear power reactors. The evaluation is performed using data from real accident histories and using representative truck and rail cask models that likely meet 10 CFR 71 regulations. The responses of the representative casks are calculated for structural and thermal loads generated by severe highway and railway accident conditions. The cask responses are compared with those responses calculated for the 10 CFR 71 hypothetical accident conditions. By comparing the responses it is determined that most highway and railway accident conditions fall within the 10 CFR 71 hypothetical accident conditions. For those accidents that have higher responses, the probabilities anf potential radiation exposures of the accidents are compared with those identified by the assessments made in the ''Final Environmental Statement on the Transportation of Radioactive Material by Air and other Modes,'' NUREG-0170. Based on this comparison, it is concluded that the radiological risks from spent fuel under severe highway and railway accident conditions as derived in this study are less than risks previously estimated in the NUREG-0170 document

  11. Summary and conclusions of the specialist meeting on severe accident management programme development

    International Nuclear Information System (INIS)

    The CSNI Specialist meeting on severe accident management programme development was held in Rome and about seventy experts from thirteen countries attended the meeting. A total of 27 papers were presented in four sessions, covering specific aspects of accident management programme development. It purposely focused on the programmatic aspects of accident management rather than on some of the more complex technical issues associated with accident management strategies. Some of the major observations and conclusions from the meeting are that severe accident management is the ultimate part of the defense in depth concept within the plant. It is function and success oriented, not event oriented, as the aim is to prevent or minimize consequences of severe accidents. There is no guarantee it will always be successful but experts agree that it can reduce the risks significantly. It has to be exercised and the importance of emergency drills has been underlined. The basic structure and major elements of accident management programmes appear to be similar among OECD member countries. Dealing with significant phenomenological uncertainties in establishing accident management programmes continues to be an important issue, especially in confirming the appropriateness of specific accident management strategies

  12. An analysis of station blackout sequences for the severe accident analysis database (II)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Dong Ha

    2006-08-15

    This report contains analysis methodologies and calculation results of station blackout sequences for the severe accident analysis database system. The Korean standard nuclear power plant has been selected as a reference plant. Based on the probabilistic safety analysis of the corresponding plant. Eight accident scenarios, which was predicted to have more than 10{sup -10}/ry occurrence frequency have been analyzed as base cases for the station blackout sequence database. Furthermore, the sensitivity studies for operational plant systems and for phenomenological models of the analysis computer code have been performed. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results of station blackout sequence in this report will be utilized as input data of the severe accident analysis database system.

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

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, Kazuichiro [ed.

    2000-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

  17. Severe Accident Progression and Consequence Assessment Methodology Upgrades in ISAAC for Wolsong CANDU6

    International Nuclear Information System (INIS)

    Amongst the applications of integrated severe accident analysis codes like ISAAC, the principal are to a) help develop an understanding of the severe accident progression and its consequences; b) support the design of mitigation measures by providing for them the state of the reactor following an accident; and c) to provide a training platform for accident management actions. After Fukushima accident there is an increased awareness of the need to implement effective and appropriate mitigation measures and empower the operators with training and understanding about severe accident progression and control opportunities. An updated code with reduced uncertainties can better serve these needs of the utility making decisions about mitigation measures and corrective actions. Optimal deployment of systems such as PARS and filtered containment venting require information on reactor transients for a number of critical parameters. Thus there is a greater consensus now for a demonstrated ability to perform accident progression and consequence assessment analyses with reduced uncertainties. Analyses must now provide source term transients that represent the best in available understanding and so meaningfully support mitigation measures. This requires removal of known simplifications and inclusion of all quantifiable and risk significant phenomena. Advances in understanding of CANDU6 severe accident progression reflected in the severe accident integrated code ROSHNI are being incorporated into ISAAC using CANDU specific component and system models developed and verified for Wolsong CANDU 6 reactors. A significant and comprehensive upgrade of core behavior models is being implemented in ISAAC to properly reflect the large variability amongst fuel channels in feeder geometry, fuel thermal powers and burnup. The paper summarizes the models that have been added and provides some results to illustrate code capabilities. ISAAC is being updated to meet the current requirements and

  18. Severe accident assessment. Results of the reactor safety research project VAHTI

    International Nuclear Information System (INIS)

    The report provides a summary of the publicly funded nuclear reactor safety research project Severe Accident Management (VAHTI). The project has been conducted at the Technical Research Centre of Finland (VTT) during the years 1994-96. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The project was divided into five work packages: (1) thermal hydraulic validation of the APROS code, (2) core melt progression within a BWR pressure vessel, (3) failure mode of the BWR pressure vessel, (4) Aerosol behaviour experiments, and (5) development of a computerized severe accident training tool

  19. 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

  20. WASA-BOSS. ATHLET-CD model for severe accident analysis for a generic KONVOI reactor

    Energy Technology Data Exchange (ETDEWEB)

    Tusheva, Polina; Schaefer, Frank; Kozmenkov, Yaroslav; Kliem, Soeren [Helmholtz-Zentrum Dresden-Rossendorf, Dresden (Germany). Reactor Safety Div.; Hollands, Thorsten [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) gGmbH, Garching (Germany); Trometer, Ailine; Buck, Michael [Stuttgart Univ. (Germany). Dept. of Reactor Safety, Systems and Environment

    2015-07-15

    Within the scope of the ongoing joint research project WASA-BOSS (Weiterentwicklung und Anwendung von Severe Accident Codes - Bewertung und Optimierung von Stoerfallmassnahmen) an ATHLET-CD model for investigation of severe accident scenarios has been developed. The model represents a generic pressurized water reactor (PWR) of type KONVOI. It has been applied for analyzing selected hypothetical core degradation scenarios, considering application of countermeasures and accident management measures, during the early phase of an accident, as well as the late in-vessel phase, when the core degradation process has already begun. Possible accident management measures for loss of coolant (LOCA) and station blackout (SBO) scenarios are discussed. This paper focuses on the ATHLET-CD model development and results from selected simulations for a SBO scenario without and with application of countermeasures.

  1. A review of iodine chemistry under severe accident conditions

    International Nuclear Information System (INIS)

    This report reviews the progress that has been made in establishing a basic understanding of the factors which will determine the behaviour of iodine during postulated accidents in water-cooled reactors. The topics considered are thermal reactions, radiolytic reactions, impurity effects, organic iodide formation, integral models and tests and volatility control. There have been substantial gains in a number of areas, most notably in the kinetics and thermodynamics databases for thermal and radiolytic reactions of inorganic iodine in solution. However, there remains a limited understanding of the mechanisms controlling the formation of organic iodides and a need for integral tests of iodine behaviour in complex, 'dirty' systems to provide data for the validation of chemical models which are undergoing development. 81 refs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  5. Simulation of a low-pressure severe accident scenario in a PWR with ATHLET-CD

    Energy Technology Data Exchange (ETDEWEB)

    Hoffmann, Mathias; Koch, Marco K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2013-07-01

    The plant behavior of a Pressurized Water Reactor (PWR) during a severe accident scenario is analyzed with system code ATHLET-CD Mod. 2.2C in order to assess the code capabilities in terms of the late-phase of the core degradation. For this purpose a severe accident sequence caused by a Station Black-out and a large break in the primary cooling system is simulated both without any accident management measures and with a delayed reflooding of the substantially degraded core. Selected code results are presented in this paper. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

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

    OpenAIRE

    MohammadJavad Jafari; Iraj Mohammadfam; Esmaeil Zarei

    2015-01-01

    Severe accidents of process industries in Iran have increased significantly in recent decade. This study quantitatively analyzes the hazards of severe accidents imposed on people, equipment and building by a hydrogen production facility. A hazard identification method was applied. Then a consequence simulation was carried out using PHAST 6.54 software package and at the end, consequence evaluation was carried out based on the best-known and different criteria. Most hazardous jet fire and flas...

  9. Impact of severe accidents on the European pressurized water reactor (ERP) design and layout

    International Nuclear Information System (INIS)

    The purpose of this presentation is to describe the impact of severe accidents on the EPR design and layout. After a summary of the safety requirements specified in accordance with the recommendations expressed by the French and German safety authorities, the main EPR features corresponding to the prevention and the mitigation of severe accidents will be described. Considerations with regard to R and D and cost impacts are also provided

  10. A space-time multivariate Bayesian model to analyse road traffic accidents by severity

    OpenAIRE

    Boulieri, A; Liverani, S; Hoogh, K. de; Blangiardo, M.

    2016-01-01

    The paper investigates the dependences between levels of severity of road traffic accidents, accounting at the same time for spatial and temporal correlations. The study analyses road traffic accidents data at ward level in England over the period 2005–2013. We include in our model multivariate spatially structured and unstructured effects to capture the dependences between severities, within a Bayesian hierarchical formulation. We also include a temporal component to capture the time effects...

  11. Review of current status for designing severe accident management support system

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwang Sub

    2000-05-01

    The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  13. Formation of decontamination cost calculation model for severe accident consequence assessment

    International Nuclear Information System (INIS)

    In previous studies, the authors developed an index “cost per severe accident” to perform a severe accident consequence assessment that can cover various kinds of accident consequences, namely health effects, economic, social and environmental impacts. Though decontamination cost was identified as a major component, it was taken into account using simple and conservative assumptions, which make it difficult to have further discussions. The decontamination cost calculation model was therefore reconsidered. 99 parameters were selected to take into account all decontamination-related issues, and the decontamination cost calculation model was formed. The distributions of all parameters were determined. A sensitivity analysis using the Morris method was performed in order to identify important parameters that have large influence on the cost per severe accident and large extent of interactions with other parameters. We identified 25 important parameters, and fixed most negligible parameters to the median of their distributions to form a simplified decontamination cost calculation model. Calculations of cost per severe accident with the full model (all parameters distributed), and with the simplified model were performed and compared. The differences of the cost per severe accident and its components were not significant, which ensure the validity of the simplified model. The simplified model is used to perform a full scope calculation of the cost per severe accident and compared with the previous study. The decontamination cost increased its importance significantly. (author)

  14. Source term estimation during incident response to severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    This document presents a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. 39 refs., 48 figs., 19 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-03-15

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

  16. Source term estimation during incident response to severe nuclear power plant accidents. Draft

    International Nuclear Information System (INIS)

    The various methods of estimating radionuclide release to the environment (source terms) as a result of an accident at a nuclear power reactor are discussed. The major factors affecting potential radionuclide releases off site (source terms) as a result of nuclear power plant accidents are described. The quantification of these factors based on plant instrumentation also is discussed. A range of accident conditions from those within the design basis to the most severe accidents possible are included in the text. A method of gross estimation of accident source terms and their consequences off site is presented. The goal is to present a method of source term estimation that reflects the current understanding of source term behavior and that can be used during an event. (author)

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

    Science.gov (United States)

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

    2016-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1993-01-15

    As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  1. Key factors contributing to accident severity rate in construction industry in Iran: a regression modelling approach.

    Science.gov (United States)

    Soltanzadeh, Ahmad; Mohammadfam, Iraj; Moghimbeigi, Abbas; Ghiasvand, Reza

    2016-03-01

    Construction industry involves the highest risk of occupational accidents and bodily injuries, which range from mild to very severe. The aim of this cross-sectional study was to identify the factors associated with accident severity rate (ASR) in the largest Iranian construction companies based on data about 500 occupational accidents recorded from 2009 to 2013. We also gathered data on safety and health risk management and training systems. Data were analysed using Pearson's chi-squared coefficient and multiple regression analysis. Median ASR (and the interquartile range) was 107.50 (57.24- 381.25). Fourteen of the 24 studied factors stood out as most affecting construction accident severity (psafety and health risk management system to reduce ASR. PMID:27092639

  2. Analysis of causes and sequences of the accident on Fukushima NPP as a factor of sever accidents prevention in the vessel reactor

    International Nuclear Information System (INIS)

    In this monograph, the provisional analysis of the causes and sequences of the sever accidents on the Fukushima NPP is presented. The analysis of the possibility of the origin of extreme events connected with the flooding of Zaporizhzhia NPP industrial site, emergency of the steam-gas explosions on NPPs with WWER and other phenomena occurred under sever accidents was carried out. It was presented the authors original working-out on symptom-oriented approaches of sever accident initiating event list identification, on criteria substantiation of explosion safety and optimization of processes management at sever accidents, as well as on the methodological support of the accident beyond the design basis management at the WWER for prevention of their transition in the stage of sever accidents.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  6. Requirement analysis of computerized procedures of AP1000 severe accident management guidelines

    International Nuclear Information System (INIS)

    Computerized procedures are drawing increased interest for application in nuclear power plants to enhance operator performance, especially in the accident conditions. AP1000 Severe Accident Management Guidelines (SAMG) are established to protect the containment fission product boundaries and to mitigate the accident consequences. This paper introduces the AP1000 SAMG, and according to the functional requirements of the Computerized Procedure System (CPS), some requirements are analyzed. These requirements are special to the Computerized AP1000 SAMG, which need to be especially noticed in the design process. (author)

  7. Cleanup and decommissioning of a nuclear reactor after a severe accident

    International Nuclear Information System (INIS)

    Although the development of commercial nuclear power plants has in general been associated with an excellent record of nuclear safety, the possibility of a severe accident resulting in major fuel and core damage cannot be excluded and such accidents have in fact already occurred. For over a decade, IAEA publications have provided technical guidance and recommendations for post-accident planning to be considered by appropriate authorities. Guidance and recommendations have recently been published on the management of damaged nuclear fuel, sealing of the reactor building and related safety and performance assessment aspects. The present technical report on the cleanup and decommissioning of reactors which have undergone a severe accident represents a further publication in the series. Refs, figs and tabs.

  8. Analysis on thermal load response for the in-vessel retention during a severe accident

    International Nuclear Information System (INIS)

    A thermal load from the molten pool in the lower plenum to the reactor vessel during a severe accident has been analyzed. The configuration of the molten pool was considered as a two-layer. A heat flux distribution, crust thickness and vessel thickness were mainly investigated in this study. Non-linear Newton-Raphson iteration method was easily applied to solve a set of governing equations. Of many severe accident sequences, SBLOCA (Small Break Loss-Of-Coolant Accident) and LBLOCA (Large Break Loss-Of-Coolant Accident) without SI (Safety Injection) in the APR1400 were considered. From the results, the focusing effect in light metallic layer could be seen and other important parameter was also explained. (author)

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-06-01

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

  13. Severe accident management (SAM), operator training and instrumentation capabilities - Summary and conclusions

    International Nuclear Information System (INIS)

    The Workshop on Operator Training for Severe Accident Management (SAM) and Instrumentation Capabilities During Severe Accidents 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 Workshop was organised into five sessions: - 1: Introduction; - 2: Tools and Methods; - 3: Training Programmes and Experience; - 4: SAM Organisation Efficiency; - 5: Instrumentation Capabilities. It was concluded by a Panel and General Discussion. This report presents the summary and conclusions: the meeting confirmed that only limited information is needed for making required decisions

  14. The Effect of the SIT Injection Rate on Delaying RPV Failure during Severe Accidents

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Seungwon; Jeun, Gyoodong; Kim, Sung Joong [Hanyang Univ., Seoul (Korea, Republic of); Kim, Hwan-Yeol; Ha, Kwang Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    Studies on reducing such complexity and uncertainty have been conducted recently. A great concern is focused on the core coolability in the case of hypothesized severe accidents. In specific, the relationship between entering severe accident management guidance (SAMG) and accompanying reactor pressure vessel (RPV) failure time was analyzed in our previous study. Lee et al. studied the effect of mitigation strategy of three postulated accidents using the MELCOR code. The objective of this study is to investigate the SAMG entry condition effect on delaying RPV failure. Delayed RPV failure time was measured for postulated scenarios with various SAMG entry conditions. Also mass balance analysis focused on reactor coolant system (RCS) is conducted to investigate safety injection tank (SIT) effect on the RPV failure. Severe accident code MELCOR 1.8.6 was used to simulate and a reference plant is selected as the Korean Pressurized Water Reactor (PWR) Optimized Pressurized Reactor (OPR) 1000. The effectiveness of SAMG entry conditions for the postulated severe accident scenarios was analyzed with the MELCOR 1.8.6. First, SBLOCA, SBO, and TLOFW with mitigation strategies were investigated in terms of delaying RPV failure. The performance of mitigation strategies with four SAMG entry conditions (the time when CET=753, 838, 923 and 973 K) was analyzed.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  16. Proceedings of the first OECD (NEA) CSNI-Specialist Meeting on Instrumentation to Manage Severe Accidents

    International Nuclear Information System (INIS)

    OECD member countries have adopted various accident management measures and procedures. To initiate these measures and control their effectiveness, information on the status of the plant and on accident symptoms is necessary. This information includes physical data (pressure, temperatures, hydrogen concentrations, etc.) but also data on the condition of components such as pumps, valves, power supplies, etc. In response to proposals made by the CSNI - PWG 4 Task Group on Containment Aspects of Severe Accident Management (CAM) and endorsed by PWG 4, CSNI has decided to sponsor a Specialist Meeting on Instrumentation to Manage Severe Accidents. The knowledge-basis for the Specialist Meeting was the paper on 'Instrumentation for Accident Management in Containment'. This technical document (NEA/CSNI/R(92)4) was prepared by the CSNI - Principle Working Group Number 4 of experts on January 1992. The Specialist Meeting was structured in the following sessions: I. Information Needs for Managing Severe Accidents, II. Capabilities and Limitations of Existing Instrumentation, III. Unconventional Use and Further Development of Instrumentation, IV. Operational Aids and Artificial Intelligence. The Specialist Meeting concentrated on existing instrumentation and its possible use under severe accident conditions; it also examined developments underway and planed. Desirable new instrumentation was discussed briefly. The interactions and discussions during the sessions were helpful to bring different perspectives to bear, thus sharpening the thinking of all. Questions were raised concerning the long-term viability of current (or added) instrumentation. It must be realized that the subject of instrumentation to manage severe accidents is very new, and that no international meeting on this topic was held previously. One of the objectives was to bring this important issue to the attention of both safety authorities and experts. It could be seen from several of the presentations and from

  17. Development and application of a radioactivity evaluation technique the to obtain radiation exposure dose of radioactivity evaluation technique when a severe accident occurs in the a power station of a severe accident. Accident management guidelines of knowledge-based maintenance

    International Nuclear Information System (INIS)

    As a One of the lessons learned from the nuclear accident at the Fukushima Daiichi Nuclear Power Stations of Tokyo Electric Power Company, the was the need for improvement of accident management guidelines is required. In this report study, we developed and applied a dose evaluation technique to evaluated the radiation dose in a nuclear power plant assuming three conditions: employees were evacuation evacuated at the time of a severe accident occurrence; operators carried out the accident management operation; of the operators, and the repair work was carried out for of the trouble damaged apparatuses in a the nuclear power plant using a dose evaluation system. The following knowledge findings were obtained and should to be reflected to in the knowledge base of the guidelines was obtained. (1) By making clearly identifying an areas beforehand becoming the that would receive high radiation doses at the time of a severe accident definitely beforehand, we can employees can be moved to the evacuation places through an areas having of low dose rate and it is also known it how much we long employees can safely stay in the evacuation places. (2) When they circulate CV containment vessel recirculation sump water is recirculated by for the accident management operation and the restoration of safety in the facilities, because the plumbing piping and the apparatuses become radioactive radioactivity sources, the dose evaluation of the shortest access route and detour access routes with should be made for effective the accident management operation is effective. Because the area where a dose rate rises changes which as safety apparatuses are restored, in consideration of a plant state, it is necessary to judge the rightness or wrongness of the work continuation from the spot radioactive dose of the actual apparatus area, with based on precedence of the need to restore with precedence, and to choose a system to be used for accident management. (author)

  18. Development of Instrument Transmitter Protecting Device against High-Temperature Condition during Severe Accidents

    Directory of Open Access Journals (Sweden)

    Min Yoo

    2014-01-01

    Full Text Available Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations, and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

  19. PCTRAN-3: The third generation of personal computer-based plant analyzer for severe accident management

    International Nuclear Information System (INIS)

    PCTRAN is a plant analyzer that uses a personal computer to simulate plant response. The plant model is recently expanded to accommodate beyond design-basis severe accidents. In the event of multiple failures of the plant safety systems, the core may experience heatup and extensive failure. Using a high-powered personal computer (PC), PCTRAN-3 is designed to operate at a speed significantly faster than real-time. A convenient, interactive and user-friendly graphics interface allows full control by the operator. The plant analyzer is intended for use in severe accident management. In this paper the code's component models and sample runs ranging from normal operational transients to severe accidents are reviewed. (author)

  20. Multi-physics particle method for the simulation of severe accidents in nuclear power plants

    International Nuclear Information System (INIS)

    The development of analysis methods for severe accidents in nuclear reactors is a key issue for nuclear safety. It is difficult to estimate the behaviors of several phenomena in reactor accidents, such as the melting and relocation of structural materials, the spreading of corium on the ground, and the molten core concrete interaction (MCCI), because they involve large changes of the geometry. In the present study, a new method was developed to simulate these phenomena by using a moving particle semi-implicit (MPS) method with models for surface tension, rigid bodies, melting and freezing, heat conduction, interfacial heat transfer, and heat radiation. As benchmarks, the melting of a metal cylinder on a hot plate and the freezing of a molten metallic drop in a coolant are simulated. The characteristic behaviors in each experiment agreed well with the simulation results, which indicates that the developed method is applicable for simulation to evaluate the behavior of corium in severe reactor accidents. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    DEFF Research Database (Denmark)

    Kaplan, Sigal; Prato, Carlo Giacomo

    2012-01-01

    that accident severity increases: (i) for young bus drivers under the age of 25; (ii) for drivers beyond the age of 55, and most prominently for drivers over 65 years old; (iii) for female drivers; (iv) for very high (over 65 mph) and very low (under 20 mph) speed limits; (v) at intersections; (vi) because...... of 2011. Method: The current study investigates the underlying risk factors of bus accident severity in the United States by estimating a generalized ordered logit model. Data for the analysis are retrieved from the General Estimates System (GES) database for the years 2005–2009. Results: Results show...

  5. Severe accident analyses for shutdown modes and spent fuel pools to support PSA level 2 activities

    Energy Technology Data Exchange (ETDEWEB)

    Kowalik, M.; Mildenberger, O.; Loeffler, H.; Steinroetter, T. [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Koeln (Germany)

    2013-07-01

    In the field of Level 2 PSA at GRS two projects are being performed in order to investigate both shutdown modes and severe accident sequences following from external hazards of nuclear power plants as well as spent fuel pool behavior under severe accident conditions. These works are being done for both PWR and BWR respectively. For both projects, deterministic severe accident analyses using the MELCOR code are a main part of the activities in order to support the probabilistic part of these projects. The German Federal Ministry for the Environment, Nature Conservation and Nuclear Safety (BMU) and the Federal Office for Radiation Protection (BfS) financially support a project regarding deterministic analyses of severe accident sequences during shutdown modes and external hazards (flooding, aircraft crash, earthquakes and explosions pressure wave). These results can be used for supporting future Level 2 PSA studies. Within a research project financially supported by the German Federal Ministry of Economics and Technology (BMWi) an extension of probabilistic analyses of spent fuel pools is being performed. Appropriate methods for the consideration to spent fuel pools inside a PSA Level 2 will be developed. The main goals are the identification of the impact of severe accidents inside spent fuel pools onto the plant behavior and the quantification of related releases of radionuclides into the environment. Results of MELCOR analyses done for the two projects mentioned above are presented. First, preliminary results of a severe accident sequence initiated by a loss of decay heat removal of a PWR shutdown mode are discussed. Following, preliminary results of the PWR spent fuel pool behavior after a 'Station Black-out' are shown. It could be shown that the integral code MELCOR is able to calculate the accident progression of an event starting from a shutdown mode of a PWR and the severe accident sequence inside of a PWR spent fuel pool. The results seem to be

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

    International Nuclear Information System (INIS)

    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

  7. Aerosol challenges to air cleaning systems during severe accidents in nuclear plants

    International Nuclear Information System (INIS)

    A variety of air cleaning systems may be operating in nuclear power plants and under severe accident conditions, these systems may be treating airborne concentrations of aerosols which are very high. Predictions of airborne aerosol concentrations in nuclear power plant containments under severe accident conditions are reviewed to provide a basis for evaluating the potential effects on the air cleaning systems. The air cleaning systems include filters, absorber beds, sprays, water pools, ice beds, and condensers. Not all of these were intended to operate as air cleaners but will in fact be good aerosol collectors. Knowledge of expected airborne concentrations will allow better evaluation of system performances

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

    International Nuclear Information System (INIS)

    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)

  9. Simulation of the Lower Head Boiling Water Reactor Vessel in a Severe Accident

    Directory of Open Access Journals (Sweden)

    Alejandro Nuñez-Carrera

    2012-01-01

    Full Text Available The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the failure of the Reactor Pressure Vessel (RPV. The release of molten material to the primary containment and the possibility of steam explosion may produce the failure of the primary containment with high radiological consequences. Then, it is important to have a detailed model in order to predict the behavior of the reactor vessel lower head in a severe accident. In this paper, a hypothetical simulation of a Loss of Coolant Accident (LOCA with simultaneous loss of off-site power and without injection of cooling water is presented with the proposal to evaluate the temperature distribution and heatup of the lower part of the RPV. The SCDAPSIM/RELAP5 3.2 code was used to build the BWR model and conduct the numerical simulation.

  10. Implementation of the severe accident management in Slovenske Elektrarne, subsidiary of ENEL

    International Nuclear Information System (INIS)

    Implementation of the Severe Accident Management (SAM) in Slovenske Elektrarne, subsidiary of ENEL, is a process initiated well before the Fukushima Daiichi accident. The main goal was to cover, comprehensively, level 4 of the Defense in Depth (DiD). The process included development of plant specific severe accident management guidelines (SAMGs) and installation of hardware modifications dedicated to mitigation of severe accidents as an upgrade the original VVER-440/V213. The SAM modifications have been developed with the aim to address all main generic vulnerabilities of VVER-440/V213 containments identified during initial analysis supporting the development of plant specific SAMGs. SAM modifications, in addition to their original purpose, improved plant response also at the level 3 of DiD. SAM modifications installed on VVER-440/V213 units in operation or under construction in Slovakia can be considered as an independent and diverse provision for the main safety functions: core subcriticality, core heat removal and confinement integrity. Basic set of SAM modifications includes independent diesel generator (DG), independent external source of borated water and containment vacuum breaker. Major contribution to safety from SAM modifications has been proved to be the implementation of in-vessel retention, hydrogen management in the containment and reliable depressurization of RCS. The complete set of SAM modifications installed incorporates dedicated SAM I and C to allow for determination and monitoring of plant status via dedicated instrumentation and control of SAM equipment installed at plants during a severe accident. SAM project including updating of SAMGs has been successfully completed on both units of Bohunice NPP and respective activities are continuing on operating units no. 1 and 2 in Mochovce with expected deadline in 2015. The basic design of Mochovce units no. 3 and 4 that are under construction has been modified to incorporate hardware changes

  11. Medical emergency planning in case of severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    This paper is an attempt to discuss a three-step-plan on medical emergency planning in case of severe accidents at nuclear power plants on the basis of own experiences in the regional area as well as on the basis of recommendations of the Federal Minister of the Interior. The medical considerations take account of the severity and extension of an accident whereby the current definitions used in nuclear engineering for accident situations are taken as basis. A comparison between obligatory and actual state is made on the possibilities of medical emergency planning, taking all capacities of staff, facilities, and equipment available in the Federal Republic of Germany into account. To assure a useful and quick utilization of the existing infra-structure as well as nation-wide uniform training of physicians and medical assistants in the field of medical emergency in case of a nuclear catastrophe, a federal law for health protection is regarded urgently necessary. (orig.)

  12. Knowledge data base for severe accident management of nuclear power plants

    International Nuclear Information System (INIS)

    For the safety enhancement of Nuclear Power Plants (NPPs), continuous efforts are very important to take in the up-to-date scientific and technical knowledge positively and to reflect them into the safety regulation. The purpose of the present study is to gather effectively the scientific and technical knowledge about the severe accident (SA) phenomena and the accident management (AM) for prevention and mitigation of SA, and to take in the experimental data by participating in the international cooperative experiments regarding the important SA phenomena and the effectiveness of AM. Based on those data and knowledge, JNES is developing and improving severe accident analysis models to maintain the SA analysis codes and the AM knowledge base for assessment of the NPPs in Japan. The activities in fiscal year 2012 are as follows; Analytical study on OECD/NEA projects such as MCCI, SERENA and SFP projects, and support in making regulation for SA. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-12-15

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

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

    Directory of Open Access Journals (Sweden)

    Jan Christian Kaiser

    2012-01-01

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

  15. WWER Technical Support Center and Training of its Staff for Severe Accident Management

    International Nuclear Information System (INIS)

    The Russian Utility organization Concern Rosenergoatom (REA) has well developed multi-level system of prevention and liquidation of emergency situations at nuclear power plants. This system covers all aspects related to beyond design accidents - from the technical support of the plant personnel to the measures for protection of the population and environment. In case a radiation dangerous situation or accident at a NPP occurred, the urgent help is being performed by the OPAS group, which coordinates the activities of forces and means participating in localization and liquidation of accident. Technical and information needs of the OPAS group is assured by Crisis center of REA (CC) with its Expert group. The task of CC is the development of the technical recommendations for the plant personnel on the accident management measures aimed to prevent the severe accident or to restrict its consequences. This task is being solved by Expert group (EG) of Crisis center in interaction with the Technical support centers (TSC) established in different design and scientific organizations (NSSS General designer, NPP General designer, Scientific leader of NPP design, institutes of Academy of Sciences, etc). Each TSC is being considered as a constituent of Rosenergoatom CC. Such Technical support center for WWER nuclear power plants (WWER TCS) has been established in OKB Gidropress some years ago. Three modes of WWER TSC operation (and, accordingly, its interaction with REA CC) are defined: normal operation, increased readiness and emergency situation. In case of beyond design accident on a plant, WWER TSC under request of REA CC will develop the recommendations for CC Expert group aimed to prevent the accident progression to the severe phase or to restrict the severe accident consequences, if it nevertheless has occurred. In chapter 2 of the present paper, place and role of WWER TSC in general system of emergency response of Rosenergoatom is highlighted. TSC structure, functions of

  16. Instrumentation Capabilities. Their Influence on Severe Accident Management and How Operator Training can be contemplated

    International Nuclear Information System (INIS)

    No currently operating nuclear unit has been explicitly designed to withstand the loads resulting from accident sequences resulting in melting of a very significant portion of the core. As a consequence, instrumentation needs were defined based on what was deemed necessary to control the unit during normal operation and contemplated accident sequences. Detailed requirements for instrumentation were then established based on environmental conditions anticipated during accident sequences addressed in the design, estimation of additional conservatism deemed reasonable for assessing sensor robustness and information reliability, and a realistic understanding of the influence of aging. Though instrument failures could not be excluded, consequences were necessarily limited as adequate redundancy was provided by design for all information needed to adequately control the unit and bring it back to safe shutdown in case of accident could be assumed available. Training programs largely built on this very robust approach and operators were challenged to control situations whose main attributes were: - all systems needed to fulfill essential safety functions are available and have the minimal capability for allowing compliance with otherwise stated acceptance criteria, - information needed to make decisions is available and reliable, - plant evolution, if not easily understandable in all cases, is not confusing to operators as all involved physical phenomena are unambiguous on one side, and can be reasonably well monitored. However, though current plant designs are generally very robust, one cannot exclude that accident sequences involving significant melting of the core can happen. First estimates through risk studies reported in WASH-1400 showed that the risk of core-melt could not be ignored, and the TMI-2 accident in a first step, then Chernobyl confirmed this conclusion. These events gave impetus to the development of Severe Accident Management (SAM) programs, and

  17. Fukushima Nuclear Accident, the Third International Severe Nuclear Power Plant Accident

    International Nuclear Information System (INIS)

    Japan is the world's third largest power user. Japan's last remaining nuclear reactor shutdown on Saturday 4 Th of May 2012 leaving the country entirely nuclear free. All of 50 of the nation's operable reactors (not counting for the four crippled reactors at Fukushima) are now offline. Before last year's Fukushima nuclear disaster, the country obtained 30% of its energy from nuclear plants, and had planned to produce up to 50% of its power from nuclear sources by 2030. Japan declared states of emergency for five nuclear reactors at two power plants after the units lost cooling ability in the aftermath of Friday 11 March 2011 powerful earthquake. Thousands of (14000) residents were immediately evacuated as workers struggled to get the reactors under control to prevent meltdowns. On March 11 Th, 2011, Japan experienced a sever earthquake resulting in the shutdown of multiple reactors. At Fukushima Daiichi site, the earthquake caused the loss of normal Ac power. In addition it appeals that the ensuing tsunami caused the loss of emergency Ac power at the site. Subsequent events caused damage to fuel and radiological releases offsite. The spent fuel problem is a wild card in the potentially catastrophic failure of Fukushima power plant. Since the Friday's 9.0 earthquake, the plant has been wracked by repeated explosions in three different reactors. Nuclear experts emphasized there are significant differences between the unfolding nuclear crisis at Fukushima and the events leading up to the Chernobyl disaster in 1986. The Chernobyl reactor exploded during a power surge while it was in operation and released a major cloud of radiation because the reactor had no containment structure around to. At Fukushima, each reactor has shutdown and is inside a 20 cm-thick steel pressure vessel that is designed to contain a meltdown. The pressure vessels themselves are surrounded by steel-lined, reinforced concrete shells. Chernobyl disaster was classified 7 on the International

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1998-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  3. The COLIMA experiment on aerosol retention in containment leak paths under severe nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Parozzi, Flavio, E-mail: flavio.parozzi@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Caracciolo, Eduardo D.J., E-mail: eduardo.caracciolo@rse-web.it [RSE, Power Generation Department, via Rubattino 54, I-20134 Milano (Italy); Journeau, Christophe, E-mail: christophe.journeau@cea.fr [CEA Cadarache (France); Piluso, Pascal, E-mail: pascal.piluso@cea.fr [CEA Cadarache (France)

    2013-08-15

    Highlights: ► Experiment investigating aerosol retention within concrete containment cracks under nuclear severe accident conditions. ► Provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. ► Prototypical aerosol particles generated with a thermite reaction and transported through the crack sample reproducing surface characteristics, temperature, pressure drop and gas leakage. ► The results indicate the significant retention due to zig-zag path. -- Abstract: CEA and RSE managed an experimental research concerning the investigation of aerosol retention within concrete containment cracks under severe accident conditions. The main experiment was carried out in November 2008 with aerosol generated from the COLIMA facility and a sample of cracked concrete with defined geometric characteristics manufactured by RSE. The facility provided representative conditions of the aerosols suspended inside the containment of PWRs under a severe accident. Prototypical aerosol particles were generated with a thermite reaction and transported through the crack sample, where surface characteristics, temperature, pressure drop and gas leakage were properly reproduced. The paper describes the approach adopted for the preparation of the cracked concrete sample and the dimensioning of the experimental apparatus, the test procedure and the measured parameters. The preliminary results, obtained from this single test, are also discussed in the light of the present knowledge about aerosol phenomena and the theoretical analyses of particle behaviour with the crack path.

  4. Stepwise integral scaling method and its application to severe accident phenomena

    Energy Technology Data Exchange (ETDEWEB)

    Ishii, M.; Zhang, G. [Purdue Univ., West Lafayette, IN (United States). School of Nuclear Engineering; No, H.C. [Korea Advanced Inst. of Science and Technology, Seoul (Korea, Republic of)

    1993-10-01

    Severe accidents in light water reactors are characterized by an occurrence of multiphase flow with complicated phase changes, chemical reaction and various bifurcation phenomena. Because of the inherent difficulties associated with full-scale testing, scaled down and simulation experiments are essential part of the severe accident analyses. However, one of the most significant shortcomings in the area is the lack of well-established and reliable scaling method and scaling criteria. In view of this, the stepwise integral scaling method is developed for severe accident analyses. This new scaling method is quite different from the conventional approach. However, its focus on dominant transport mechanisms and use of the integral response of the system make this method relatively simple to apply to very complicated multi-phase flow problems. In order to demonstrate its applicability and usefulness, three case studies have been made. The phenomena considered are (1) corium dispersion in DCH, (2) corium spreading in BWR MARK-I containment, and (3) incore boil-off and heating process. The results of these studies clearly indicate the effectiveness of their stepwise integral scaling method. Such a simple and systematic scaling method has not been previously available to severe accident analyses.

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

    International Nuclear Information System (INIS)

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

  6. Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report

    International Nuclear Information System (INIS)

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  8. Calculations of severe accident progression in the General Electric Simplified Boiling Water Reactor

    International Nuclear Information System (INIS)

    General Electric is designing a new nuclear power plant: the Simplified Boiling Water Reactor (SBWR). The SBWR is a passive plant in which the core cooling and decay heat removal safety systems are driven by gravity. To model the plant response to severe accidents, MAAP-SBWR, an advanced version of the Modular Accident Analysis Program (MAAP), has been developed. The main feature of the new code is a flexible containment model. The challenges in modeling the SBWR, the code structure and models, and a sample application to the SBWR are discussed

  9. Hydrogen-control systems for severe LWR accident conditions - a state-of-technology report

    Energy Technology Data Exchange (ETDEWEB)

    Hilliard, R K; Postma, A K; Jeppson, D W

    1983-03-01

    This report reviews the current state of technology regarding hydrogen safety issues in light water reactor plants. Topics considered in this report relate to control systems and include combustion prevention, controlled combustion, minimization of combustion effects, combination of control concepts, and post-accident disposal. A companion report addresses hydrogen generation, distribution, and combustion. The objectives of the study were to identify the key safety issues related to hydrogen produced under severe accident conditions, to describe the state of technology for each issue, and to point out ongoing programs aimed at resolving the open issues.

  10. Application of COREMELT-3D code at analysis of severe fast reactor accidents

    International Nuclear Information System (INIS)

    The code COREMELT for calculations of initial and transition stages of severe accident is considered. It is used to conduct connected calculations of nonstationary neutronic and thermohydraulic processes in sodium fast reactor core. The code has some versions depending on dimensions of solving problem and consists of thermohydraulic module COREMELT and neutronic module RADAR. Using the code COREMELT-3D connected calculations of core disassembly accidents of ULOF and UTOP type have been conducted for sodium fast reactors safety analysis. The main problem of code COREMELT-3D use is duration of calculation, speeding of the code is possible when calculating algorithms are parallelized

  11. Multi-step approach to Code-coupling for progression induced severe accidents in CANDU NPPs (MACPISA-CANDU)

    Energy Technology Data Exchange (ETDEWEB)

    Pohl, D.J.; Luxat, J.C. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada); Giannotti, W.; D' Auria, F. [Univ. of Pisa, Dept. of Mechanical, Nuclear and Production Engineering, Pisa (Italy)

    2009-07-01

    This paper reviews the progression of severe accidents, describes computer codes currently employed for analysis of severe accidents and outlines a new methodology to modelling the progression of severe accidents in CANDU nuclear power plants (NPPs) called the Multi-step Approach to Code-coupling for Progression Induced Severe Accidents in CANDU NPPs (MACPISA-CANDU). The MACPISA-CANDU methodology was used to couple the U.S. NRC codes SCDAP/RELAP5 (RELAP/SCDAPSIM Mod 3.4) and MELCOR (1.8.5) in order to model a small break loss of coolant accident with loss of emergency coolant injection (SBLOCA-LOECI) under natural circulation in a CANDU 6 NPP. Using this model it was shown that the sheath temperature did not exceed the zirconium melting temperature of 2098 K and hence the progression of the severe accident was terminated as expected. (author)

  12. AP-1000 passive in-vessel retention design for a severe accident

    International Nuclear Information System (INIS)

    The AP-1000 is a Westinghouse two-loop 1100 MWe advanced pressurized water reactor (PWR) that uses passive safety features to enhance plant safety and to provide significant and measurable improvements in plant simplification, reliability, investment protection and plant costs. One of the passive safety features of the AP-1000 is In-Vessel Retention. In-Vessel Retention passively provides sufficient external cooling of the reactor vessel to retain a molten core inside the vessel in the unlikely event of a severe accident. This concept was proven by a series of tests and analyses, and offers numerous advantages over other severe accident core management designs. The testing that proved the In-Vessel Retention Concept for AP-1000 also identified a series of features and functions for the reactor vessel insulation, making it different from the reactor vessel insulation in any other nuclear power plant. This paper presents the key functional and interface requirements for the AP-1000 reactor vessel insulation and the design configuration that meets those requirements. Requirements are identified for both normal and severe accident conditions. The requirements for normal conditions are similar to those for any other reactor vessel insulation. The severe accident requirements include features that passively allow containment flood water in the reactor cavity to freely flow between the insulation and reactor vessel and other features that passively allow the resulting steam to freely vent. The AP-1000 reactor vessel insulation must also provide the appropriate annulus to optimally cool the reactor vessel and be structurally adequate to withstand the loads that occur during the severe accident and during certain pipe breaks in containment. (authors)

  13. Modelling and analysis of the behavior of LWRs at severe core accidents

    International Nuclear Information System (INIS)

    With respect to the assessment of the consequences of severe accidents in light water reactors from the initiation of the accident up to the thermal failure of the reactor pressure vessel (RPV), a modular program system has been developed. Experimental results will be considered with respect to the modeling of the fuel rod behavior, e.g. deformation of the fuel rod, metal water reaction and the melting of the fuel rods. The fuel and core models allow to estimate the coolability of fuel rods and core as well as the consequences of core meltdown accidents at various pressure levels. After partial failure of the lower core retention structure, the core material will drop into the lower plenum and heat up the RPV. This strong interaction between the thermal behavior of the remaining core and the partially dropped core material has been modeled because of an accident sequence analysis. The analyses described here show, that not the entire core will fail, but a partial drop of core material into the lower plenum is likely to occur. With respect to the validation of the program system, comparison calculations with the fuel rod behavior and melt models SSYST and EXMEL will be performed. Moreover, the program system will be applied to the bundle behavior in meltdown experiments, the TMI-2 core behavior and the course of a core meltdown accident in risk studies. (orig.)

  14. Strategy of severe accident physical modeling in view of recent requirements to safety analysis

    International Nuclear Information System (INIS)

    Nuclear power destiny in various states including Russia is not free from questions. Where there is plenty of non-expensive natural gas or coal in a country, the competition of nuclear power with other power sources is especially intense. Until one considers the economic efficiency or environmental impact of the normally operating plant, the estimate of the proponents favorite choice may be rather optimistic in many cases. As soon as safety aspects of nuclear power are concerned it is necessary to answer very significant questions about the dangers resulting from severe accidents. TMI and, to a greater extent, Chernobyl, demonstrated the other aspect of the severe accident problem. It serves no purpose to dwell upon the inadequate reaction of the population on the radiation problem. It is of little use to try to prove that the health consequences of the Chernobyl or some other radiation accident are substantially overestimated. To make an advance one must substantially reduce the severe accident risk. Besides that is is necessary to give a convincing proof that such a reduction has really been made

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-04-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Darrell L. Knudson

    2014-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Joy L. Rempe; Darrell L. Knudson

    2013-03-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2013-01-01

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

  20. Highly Reliable Power and Communication System for Essential Instruments under a Severe Accident of NPPs

    International Nuclear Information System (INIS)

    In this paper, three survivable strategies to overcome the problems listed above are proposed for the essential instruments under the severe accident of NPPs. First, wire/wireless multi power systems are adopted to the essential instruments for continuous power supply. Second, wire/wireless communication systems are proposed for reliable transmission of measuring information among instruments and operators. Third, a physical protection system such as a harness and a heat isolation box is introduced to ensure operable conditions for the proposed systems. In this paper, a highly reliable strategy, which consists of wire/wireless multi power and communication systems and physical protection system is proposed to ensure the survival of the essential instruments under harsh external conditions. The wire/wireless multi power and communication systems are designed to transfer power and data in spite of the failure of conventional wired systems. The physical protection system provides operable environments to the instruments. Therefore, the proposed system can be considered as a candidate of practical and urgent remedy for NPPs under the severe accident. After the Fukushima nuclear accident, survivability of essential instruments has been emphasized for immediate and accurate response. The essential instruments can measure environment conditions such as temperature, pressure, radioactivity and corium behavior inside nuclear power plants (NPPs) under a severe accident. Access to the inside of NPPs is restricted to human beings because of hazardous environment such as high radioactivity, high temperature and high pressure. Thus, monitoring the inside of NPPs is necessary for avoiding damage from the severe accident. Even though there were a number of instruments in Fukushima Daiichi NPP, they failed to obtain exact monitoring information. According to the details of the Fukushima nuclear accident, following problems can be counted as strong candidates of this instruments

  1. Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  2. Occupational Radiation Protection in Severe Accident Management. EG-SAM Interim Report

    International Nuclear Information System (INIS)

    As an early response to the Fukushima NPP accident, the 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 actors 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 protection job coverage during severe accident initial response and recovery

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

    Energy Technology Data Exchange (ETDEWEB)

    Sebrell, W.

    1983-07-01

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

  4. Use of a fuzzy decision-making method in evaluating severe accident management strategies

    International Nuclear Information System (INIS)

    In developing severe accident management strategies, an engineering decision would be made based on the available data and information that are vague, imprecise and uncertain by nature. These sorts of vagueness and uncertainty are due to lack of knowledge for the severe accident sequences of interest. The fuzzy set theory offers a possibility of handling these sorts of data and information. In this paper, the possibility to apply the decision-making method based on fuzzy set theory to the evaluation of the accident management strategies at a nuclear power plant is scrutinized. The fuzzy decision-making method uses linguistic variables and fuzzy numbers to represent the decision-maker's subjective assessments for the decision alternatives according to the decision criteria. The fuzzy mean operator is used to aggregate the decision-maker's subjective assessments, while the total integral value method is used to rank the decision alternatives. As a case study, the proposed method is applied to evaluating the accident management strategies at a nuclear power plant

  5. Proceedings of the second OECD specialist meeting on operator aids for severe accident management - SAMOA-2

    International Nuclear Information System (INIS)

    The second OECD Specialist Meeting on Operator Aids for Severe Accident Management (SAMOA-2) was organized in Lyon, France from 8 to 10 September 1997 in collaboration with the Thermal and Nuclear Studies and Project Department (SEPTEN) of Electricite de France. It was attended by 33 specialists representing ten OECD Member countries, the OECD Halden Reactor Project, the Commission of the European Communities, and the Russian Federation. The scope of SAMOA-2 was limited to operator aids for accident management which were in operation or could be soon. The meeting concentrated on the management of accidents beyond the design basis, including tools which might be extended from the design basis range into the severe accident area. Relevant simulation tools for operator training were also part of the scope of the meeting. Twenty papers were presented during the meeting, grouped into three sessions. Session 1: operator aids for control rooms; Session 2: operator aids for technical support centres; session 3: simulation tools for operator training. There were two demonstrations of computerized systems: the ATLAS analysis simulator developed by GRS, and EDF's 'Simulateurs Post Accidentels' (SIPA). There was also a video demonstration of the Full Scope Simulator developed by a joint Russian-U.S. team for the Leningrad nuclear power

  6. Uncertainty Analysis of the Potential Hazard of MCCI during Severe Accidents for the CANDU6 Plant

    Directory of Open Access Journals (Sweden)

    Sooyong Park

    2015-01-01

    Full Text Available This paper illustrates the application of a severe accident analysis computer program to the uncertainty analysis of molten corium-concrete interaction (MCCI phenomena in cases of severe accidents in CANDU6 type plant. The potential hazard of MCCI is a failure of the reactor building owing to the possibility of a calandria vault floor melt-through even though the containment filtered vent system is operated. Meanwhile, the MCCI still has large uncertainties in several phenomena such as a melt spreading area and the extent of water ingression into a continuous debris layer. The purpose of this study is to evaluate the MCCI in the calandria vault floor via an uncertainty analysis using the ISAAC program for the CANDU6.

  7. Review of five investigation committees' reports on the Fukushima Dai-ichi Nuclear Power Plant severe accident. Focusing on accident progression and causes

    International Nuclear Information System (INIS)

    On March 11, 2011, the Tohoku District-off the Pacific Ocean Earthquake and the subsequent tsunami resulted in the severe core damage at TEPCO's Fukushima Dai-ichi Nuclear Power Station Units 1-3, involving hydrogen explosions at Units 1, 3, and 4 and the large release of radioactive materials to the environment. Four independent committees were established by the Japanese government, the Diet of Japan, the Rebuild Japan Initiative Foundation, and TEPCO to investigate the accident and published their respective reports. Also, the Nuclear and Industrial Safety Agency carried out an analysis of accident causes to obtain the lessons learned from the accident and made its report public. This article reviews the reports and clarifies the differences in their positions, from the technological point of view, focusing on the accident progression and causes. Moreover, the undiscussed issues are identified to provide insights useful for the near-term regulatory activities including accident investigation by the Nuclear Regulation Authority. (author)

  8. Neutronics aspects associated to the prevention and mitigation of severe accidents in sodium cooled reactor cores

    International Nuclear Information System (INIS)

    Among all the types of accidents to be considered for the safety licensing of a plant, some have a very low probability of occurrence but might have very important consequences: the severe accidents or Hypothetical Core Disruptive Accidents (HCDA). The studies on the scenario of these accidents are performed in parallel to the prevention studies. In this PhD report, two representative safety cases are studied: the Unprotected Loss Of Flow (ULOF) and the Total Instantaneous Blockage (TIB). The objectives are to understand what causes the reactivity increase during these accidents and to find means to reduce the energetic release of the scenario (ULOF) or to find ways to trigger the core prior to the propagation of the accident (TIB). At first, the accidents are studied in static calculations with the ERANOS code system. The accidents are divided into several steps and the reactivity insertions at each step are explained. This study shows the importance of the removal of the structures as well as of the radial leakage changes during the core slumping-down. The study also gives the amounts of fuel to be ejected or of absorber to be injected in both accidents. These values give tracks to the following more accurate studies, the transient studies. The transient studies were performed with the SIMMER code system, coupling thermo-hydraulics and neutronics. SIMMER data and algorithms have been improved so as to better predict ERANOS results (former discrepancies were up to 1.5$). The SIMMER reactivity calculation is improved by 0.8$ with variations of reactivity due to the motion of materials correctly predicted. A new algorithm for the β-effective was implemented in SIMMER so as to be more accurate and easier to manage. SIMMER is then used to calculate the secondary phase of the ULOF, while the primary phase is calculated with ERANOS thanks to some assumptions. The assumptions are very much based on the fact that the movement of materials stops whenever the energy

  9. Phenomenology and Course of Severe Accidents in PWR-Plants - Training by Teaching and Demonstration

    International Nuclear Information System (INIS)

    A special one day training course on 'Phenomenology and Course of Severe Accidents in PWR-Plants' was developed at GRS initiated by the interest of German utilities. The work was done in the frame of projects sponsored by the German Ministries for Environment, Nature Conservation and Nuclear Safety (BMU) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course will be discussed and selected parts of the training course will be presented. Demonstrations are made within this training course with the GRS simulator system ATLAS to achieve a broader understanding of the phenomena discussed and the propagation of severe accidents on a plant specific basis. The GRS simulator system ATLAS is linked in this case to the integral code MELCOR and pre-calculated plant specific severe accident calculations are used for the demonstration together with special graphics showing plant specific details. Several training courses have been held since the first one in November, 1996 especially to operators, shift personal and the management board of a German PWR. In the meantime the training course was updated and suggestions for improvements from the participants were included. In the future this type of the training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (authors)

  10. Risk factors affecting fatal bus accident severity: Their impact on different types of bus drivers.

    Science.gov (United States)

    Feng, Shumin; Li, Zhenning; Ci, Yusheng; Zhang, Guohui

    2016-01-01

    While the bus is generally considered to be a relatively safe means of transportation, the property losses and casualties caused by bus accidents, especially fatal ones, are far from negligible. The reasons for a driver to incur fatalities are different in each case, and it is essential to discover the underlying risk factors of bus fatality severity for different types of drivers in order to improve bus safety. The current study investigates the underlying risk factors of fatal bus accident severity to different types of drivers in the U.S. by estimating an ordered logistic model. Data for the analysis are retrieved from the Buses Involved in Fatal Accidents (BIFA) database from the USA for the years 2006-2010. Accidents are divided into three levels by counting their equivalent fatalities, and the drivers are classified into three clusters by the K-means cluster analysis. The analysis shows that some risk factors have the same impact on different types of drivers, they are: (a) season; (b) day of week; (c) time period; (d) number of vehicles involved; (e) land use; (f) manner of collision; (g) speed limit; (h) snow or ice surface condition; (i) school bus; (j) bus type and seating capacity; (k) driver's age; (l) driver's gender; (m) risky behaviors; and (n) restraint system. Results also show that some risk factors only have impact on the "young and elder drivers with history of traffic violations", they are: (a) section type; (b) number of lanes per direction; (c) roadway profile; (d) wet road surface; and (e) cyclist-bus accident. Notably, history of traffic violations has different impact on different types of bus drivers. PMID:26513334

  11. Examination of some assumed severe reactor accidents at the Olkiluoto nuclear power plant

    International Nuclear Information System (INIS)

    Knowledge and analysis methods of severe accidents at nuclear power plants and of subsequent response of primary system and containment have been developed in last few years to the extent that realistic source tems of the specified accident sequences can be calculated for the Finnish nuclear power plants. The objective of this investigation was to calculate the source terms of off-site consequences brought about by some selected severe accident sequences initiated by the total loss of on-site and off-site AC power at the Olkiluoto nuclear power plant. The results describing the estimated off-site health risks are expressed as conditional assuming that the accident has taken place, because the probabilities of the occurence of the accident sequences considered have not been analysed in this study. The range and probabilities of occurence of health detriments are considered by calculating consequences in different weeather conditions and taking into account the annual frequency of each weather condition and statistical population distribution. The calculational results indicate that the reactor building provides and additional holdup and deposition of radioactive substance (except coble gases) released from the containment. Furthermore, the release fractions of the core inventory to the environment of volatile fission products such as iodine, cesium and tellurium remain under 0.03. No early health effects are predicted for the surrounding population in case the assumed short-tem countermeasures are performed effectively. Acute health effects are extremely improbable even without any active countermeasure. By reducing the long-term exposure from contaminated agricultural products, the collective dose from natural long-term background radiation, for instance in the sector of 30 degrees towards the southern Finland up to the distance of 300 kilometers, would be expected to increase with 2-20 percent depending on the release considered

  12. Application of Core Exit Temperature for Effective Safety Injection Strategy of Severe Accident Management Guidance

    International Nuclear Information System (INIS)

    Due to limited time for operator's action under the postulated severe accident, immediate and short term actions are needed and relevant strategies are constructed in the SAMG. Therefore, the SAMG includes a variety of information to assist the proper operator actions. Among these, pre-calculated graphs and formulas facilitate understanding of plant status and operator's action needed for accident mitigation. These are essential for ease of application and regarded as Computational Aids (CA). The representative example is the estimation of injection flow rates for removing decay heat and oxidation heat of core, and hydrogen generation rate, to mention a few. Most of all, calculation of the necessary injection flow rate is important in order to mitigate and/or terminate core damages. In estimating the flow rate for accident mitigation, Core Exit Temperature (CET) is utilized as a key variable. CET is considered most effective and reliable means for diagnosing core state. As such, CET has been adopted as a criterion transitioning from EOPs to SAMG. In this study, the necessary flow rate is calculated utilizing simple model with CET for RCS injection in mitigation strategy of SAMG. MELCOR simulation results are introduced for the calculation. A simple model of flow rate necessary for core heat removal is developed using CET data obtained from MELCOR simulations of OPR1000. The suggested model is expected to contribute on judging the core state in its coolability and required flow injection due to ease of application. More detailed analyses are needed to normalize by including additional accident scenarios

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-01

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

  14. Fission products behaviour in UO2 submitted to nuclear severe accident conditions

    Science.gov (United States)

    Geiger, E.; Bès, R.; Martin, P.; Pontillon, Y.; Solari, P. L.; Salome, M.

    2016-05-01

    The objective of this work was to study the molybdenum chemistry in UO2 based materials, known as SIMFUELS. These materials could be used as an alternative to irradiated nuclear fuels in the study of fission products behaviour during a nuclear severe accident. UO2 samples doped with 12 stable isotopes of fission products were submitted to annealing tests in conditions representative to intermediate steps of severe accidents. Samples were characterized by SEM-EDS and XAS. It was found that Mo chemistry seems to be more complex than what is normally estimated by thermodynamic calculations: XAS spectra indicate the presence of Mo species such as metallic Mo, MoO2, MoO3 and Cs2MoO4.

  15. Safety concept of severe accident free HTR and its demonstration by the HTTR

    International Nuclear Information System (INIS)

    A severe accident free HTR (SFHTR) can achieve a higher degree of safety than an HTGR designed previously. The SFHTR is completely free from severe accidents that cause a large amount of fuel failure. Moreover, the safety concept of the SFHTR will be demonstrated by using High Temperature Engineering Test Reactor (HTTR). In the HTTR, nearly full-scale worst events postulated in the SFHTR will be tested. Demonstrable safety is essential for future nuclear reactors and is only achieved due to inherent safety features in the HTGR. Demonstrable safety and license by test are completely different approaches from existing reactors. This paper describes the safety concept of the SFHTR and safety demonstration tests of the HTTR

  16. Analysis of potential for jet-impingement erosion from leaking steam generator tubes during severe accidents.

    Energy Technology Data Exchange (ETDEWEB)

    Majumdar, S.; Diercks, D. R.; Shack, W. J.; Energy Technology

    2002-05-01

    This report summarizes analytical evaluation of crack-opening areas and leak rates of superheated steam through flaws in steam generator tubes and erosion of neighboring tubes due to jet impingement of superheated steam with entrained particles from core debris created during severe accidents. An analytical model for calculating crack-opening area as a function of time and temperature was validated with tests on tubes with machined flaws. A three-dimensional computational fluid dynamics code was used to calculate the jet velocity impinging on neighboring tubes as a function of tube spacing and crack-opening area. Erosion tests were conducted in a high-temperature, high-velocity erosion rig at the University of Cincinnati, using micrometer-sized nickel particles mixed in with high-temperature gas from a burner. The erosion results, together with analytical models, were used to estimate the erosive effects of superheated steam with entrained aerosols from the core during severe accidents.

  17. A Literature Review on Ruthenium Behaviour in Nuclear Power Plant Severe Accidents

    OpenAIRE

    Mun, Christian; Cantrel, Laurent; Madic, Charles

    2007-01-01

    During an hypothetical severe accident on a Pressurized Water Reactor (PWR), Fission-Products (FPs) are released from the nuclear fuel and may reach the reactor containment building. Among the FPs, ruthenium is of particular interest due to its ability to form volatile oxide compounds in highly oxidizing conditions. In addition, Ru is a very hazardous compound because it is chemically toxic and also because of its radiotoxicity. The topic of ruthenium is examined in terms of nuclear safety is...

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

    OpenAIRE

    Soontorn Boonta; Anchalee Sattayatham; Pairote Sattayatham

    2013-01-01

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

  19. A generic approach for containment success criteria under severe accident loads

    International Nuclear Information System (INIS)

    The U.S. Department of Energy (DOE), Office of New Production Reactors (NP), has identified safety as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR). The DOE-NP has issued the Deterministic Severe Accident Criteria (DSACs) to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements that are qualitative and quantitative bases for calculating associated loadings and containment response to those loadings, and (2) Success Criteria that specify acceptable containment response measures and limits for each problem statement. This paper is limited to a discussion of a generic approach for containment success criteria. The main elements of these success criteria are expressed in terms of elastic stresses and inelastic strains. Containment performance is based on the best estimate of failure as predicted by either stress or strain, buckling, displacements, or ability to withstand missile perforation. Since these limits are best estimates of failure, no conservatism exists in these success criteria. Rather, conservatism is to be provided in the problem statements, i.e., the quantified severe accident loads. These success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements

  20. Feasibility of shock attenuation by simple obstacles in mitigating severe accident explosive loads

    International Nuclear Information System (INIS)

    Highlights: • Effects of obstacles on shock attenuation are analyzed using the fully compressible FLUENT code. • Computations for a shock tube is validated against existing experimental data. • Computations for multiple obstacles show reduction of peak pressures. • Plant scale computations show that peak pressures can be potentially reduced by about 30–50%. • Simple obstacles are candidate shock attenuators under nuclear power plant severe accidents. - Abstract: Fukushima severe accident showed that explosions are no longer exclusive in nuclear power plants and creative safety features would be beneficial to preserve the integrity of vital components against such explosions. Based on this motivation, simple obstacles are suggested as a means to attenuate shock waves that may arise from severe accident explosions and the effectiveness is investigated by computations employing fully compressible viscous Navier–Stokes equations of the FLUENT code. Two- and three-dimensional computations for a shock tube filled with air are performed and the computed peak pressure and arrival time are validated against existing experimental data and analyses. Effects of structural shapes and rectangular rigid obstacles on shock attenuation in simple geometries are preliminarily analyzed and the results show that the end wall peak pressure is reduced by about 18–30%. As a demonstration in a plant scale, a high pressure source of 0.2 m radius semicircle is positioned at the bottom center of the two-dimensional hypothetical reactor cavity model (5.4 m by 9.4 m) and multiple rigid obstacles are positioned at the half-way between a reactor vessel and the pressure source. Incident circular shock wave is dispersed by the obstacles and the peak pressures at the reactor vessel bottom head center and the 45° regions are reduced by about 35% and 50%, respectively. This study implies the feasibility of simple obstacles for attenuating shock waves from the explosions expected under

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

    Directory of Open Access Journals (Sweden)

    Soontorn Boonta

    2013-01-01

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

  2. Study on probability of failure for RPV nozzle region under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Il Soon; Oh, Young Jin; Sim, Sang Hoon [Seoul National University, Seoul (Korea)

    2002-04-01

    Most of previous studies for creep rupture of RPV lower head under severe accident condition, have been focused on global failure of RPV lower head. In contrast, the local failure of the RPV nozzle region has not been studied in detail. This study focused the nozzle failure analysis into creep rupture evaluation of RPV lower head under severe accident condition, and this will help improve the safety assessment of nuclear power plants under severe accident conditions. The existence and features of nozzle failure in LAVA-ICI tested vessel of Korea Atomic Energy Research Institute and LHF-4 tested vessel of SNL, are examined. To understand the basic mechanical properties of nozzle material and weld metal, the tensile tests in various temperature levels and the creep rupture tests in various temperature and stress levels, are accomplished. The stress and deformation of LAVA-ICI experiments are analysed using measured basic mechanical properties. The failure time of Advanced Power Reactor 1400 (APR1400) in nozzle region was calculated using modified TMI-2 VIP model. Nozzle region failure characteristics was studied for SNL-LHF-4 experimental case using Finite Element Method (FEM). Using characteristics of nozzle failure, a new failure prediction experimental method was proposed for RPV nozzle failure. 19 refs., 43 figs. (Author)

  3. Development of evaluation method for economic consequences of severe accident at NPP

    International Nuclear Information System (INIS)

    A nuclear power plant accident has onsite and/or offsite consequences. A common framework into which many of the consequences of an accident may be translated is their economic cost. Though there are some consequence analysis codes, these are limited only to estimate only offsite consequence. In this study, economic losses are estimated for the case of onsite and offsite consequences for the sample PWR plant at Yong-Gwang site. For the estimation of offsite consequence, economic database unique to Korean economic structure is utilized as many as possible. By grouping of various cost components, each cost groups are compared each other. For the detailed estimation of offsite decontamination cost, offsite surface around the plant is divided into five types of surface. This division of surface types is agricultural field, wooded land, roof surface for representing house, asphalt road and water surface. But on the water surface, it is assumed that no decontamination operation is required. The comparison shows that at the severe accident, offsite consequences are more severe than onsite consequences. And in agricultural growing season, the consequence even becomes more severe. It also turned out that the importance and selection of much appropriate criteria will play a major role, because the economic consequences are widely varying depending on how the criteria was chosen

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-10-15

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

  5. Knowledge-based modeling of operator response for severe-accident analysis

    International Nuclear Information System (INIS)

    Studies of severe accidents in light water reactors have shown that operator response can play a crucial role in the predicted outcomes of dominant accident scenarios. Although computer codes such as MAAP are available to predict the thermal-hydraulic response, substantial knowledge about plant practices and procedures is needed to make reasonable assumptions about operator response. Based on the thermal-hydraulic state of the plant, symptom-oriented procedures provide general guidance to the operators, who then take one of several possible actions. The paper pictures this process as a feedback loop that relies heavily on the judgment of the individual safety analyst. The ability to more explicitly model the procedural guidance and operator response can help close this analytical loop and improve the overall integration and consistency of severe accident analysis. An object-oriented model for operator response characteristics and symptom-oriented procedures was developed using the NEXPERT OBJECT expert system shell. This prototype system reads MAAP transient output files and determines the instructions and operator response characteristics that are implied by the observable plant variables. A limited set of boiling water reactor (BWR6) emergency operating procedures (EOPs) was formulated as a rule set, and pattern-matching techniques were used to generate message queues for display and reports

  6. EC Research Contribution to Decision-making Processes Relevant to Severe Accident Management

    International Nuclear Information System (INIS)

    As a result of the two well-known civil nuclear accidents and of the consequent increase in safety requirements, the need to properly assess severe accident (SA) scenarios for present and future nuclear power plants (going beyond the traditional three-level defence-in-depth strategy) became evident. In this line, various research activities were launched and are performed within the Euratom Framework Programmes, in particular the completed Fourth one (F P-4, 1994-1998) and the present Fifth one (FP-5, 1998-2002). The initial orientation of the EC research activities was mainly focused on improving the understanding of the phenomena and mechanisms involved in such accidents, in order to contribute to prevent possible final radioactivity releases. A consensus on how to model those SA phenomena in accident safety analyses by means of specific tools (SA codes developed, verified and validated through experimental results provided) is reasonably advanced. Currently, the EC research activities related to severe accidents are balanced between a twofold approach aimed at assessing the risks related with severe accident scenarios and to support the development of severe accident management (SAM) strategies, together with the optimisation of backfitting measures for existing reactors or specific designs for future nuclear power plants. This new orientation is confronting difficulties, inherent to the phenomenological character of several research activities, which make a direct application of the results into SAM measures premature in some cases. In this regard, this paper presents a series of ten selected FP-5 projects with emphasis placed on the applicability of research results towards SAM strategies to be used by decision-makers amongst utilities, the nuclear industry in particular designers, and regulators. The majority of them also contain -further to the SAM approach- supporting elements focused on risk assessment. The revised programme of the key action 'Nuclear

  7. Most likely vessel lower head failure location during severe accident conditions

    International Nuclear Information System (INIS)

    The Nuclear Regulatory Commission is sponsoring a lower vessel head research program to investigate plausible modes of reactor vessel failure to determine: (a) which modes have the greatest likelihood of occurrence during a severe accident and (b) the range of core debris and accident conditions that lead to these failures. All major types of US light water reactor vessels are being considered, and both high- and low-pressure conditions are being addressed for each reactor type. The research program includes analytical and finite element calculations. In addition, high temperature creep and tensile data for predicting vessel structural response were obtained. Calculational results used to predict which failure location is more likely in a particular reactor design during a severe accident are described within this paper. Detailed analyses are being performed to investigate the relative likelihood of a BWR penetration and vessel to fail during a wide range of severe accident conditions. The analyses include applying a numerical model to obtain the penetration and vessel thermal response and applying an analytical model to investigate the relative likelihood of tube rupture and global vessel failure. Sensitivity studies considered the impact of assumptions related to debris composition, debris porosity, corium decay heat, vessel coolant mass, heat removal from the vessel, melt relocation time, and melt relocation distance on vessel and penetration response. In addition, analytically developed failure maps, which were developed in terms of dimensionless groups, are applied to extrapolate numerically-obtained results to geometries and materials occurring in PWR penetration/vessel configurations and a wide range of debris conditions

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-04-15

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

  9. Severe accident source terms for a sodium-cooled fast reactor

    International Nuclear Information System (INIS)

    Highlights: • This study analyzes offsite doses for characteristic SFR scenarios. • Models to calculate the source term for an SFR were developed for this work. • Environmental releases are small due to effectiveness of retention mechanisms. • NRC’s Quantitative Health Objectives are satisfied with high margins. - Abstract: In order to support the demonstration of a risk-informed approach to the design optimization of a sodium-cooled fast reactor (SFR), it was necessary to make realistic estimates of the consequences of severe accident scenarios. This paper describes the database, models, and assumptions used to estimate the offsite consequences of characteristic severe accident scenarios. As required for comparison with the NRC’s technology neutral framework limit curve, the offsite dose at one mile from the plant boundary is calculated using conservative meteorology. The reference plant design is a 1000 MWt pool-type design with metallic fuel. Because an integrated analysis tool comparable to MELCOR does not exist for SFR accident scenario analysis, it was necessary to write a computer code that would assess release of radionuclides from the fuel and transport within the reactor primary system and to link those analyses with results from existing computer codes that assess the dynamic response of the reactor, containment thermal–hydraulics, and radionuclide transport processes within the containment. The analyses indicate that the offsite source terms for SFR severe accident scenarios tend to be small because of the low melting temperature of the fuel, likelihood of significant retention of fission products within the sodium pool, augmentation of containment deposition processes by interaction with sodium oxide aerosols, and small driving force for release from the containment to the environment. A number of major sources of modeling uncertainty are identified as requiring further development effort. An integrated modeling capability, similar to the

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  11. Proceedings of the workshop on the implementation of severe accident management measures

    International Nuclear Information System (INIS)

    The OECD/NEA Workshop on the Implementation of Severe Accident Management (SAM) Measures was hosted by the PSI (Paul Schemer Institut), by two Swiss Utilities (Kernkraftwerk Beznau and Kernkraftwerk Leibstadt), and by Electricite de France. Eighty specialists from fourteen OECD Member countries attended the meeting, as well as specialists from three non-Member economies and the European Commission. Thirty-three papers were presented in four sessions, preceded by a brief Introductory Session (two invited papers) and followed by a General Discussion. 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

  12. Analysis of the containment spray effect for severe accident management during Molten Core-Concrete Interaction

    International Nuclear Information System (INIS)

    Massive combustible gases generated by MCCI during a severe accident in NPP causes a problem of when we should spray the containment. The increase of hydrogen concentration due to the steam condensation caused by spraying might lead to a hydrogen burning and thus intimidate the containment integrity. In case the containment is designed to be robust enough to sustain the AICC (Adiabatic Isochoric Complete Combustion) load and to prevent DDT (Deflagration to Detonation Transition), it might be effective to spray and thus burn the hydrogen at early phase of MCCI to keep the containment integrity. Spraying the containment at late phase of MCCI might cause the containment to fail because of the increased combustible gases generation. MELCOR analysis for APR1400 shows that spraying the containment at early phase can delay the time to reach containment failure pressure by steam inerting and oxygen depletion. This kind of analysis helps us to better establish a spray actuation time for an accident management procedure against a postulated severe accident

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. Prediction of Reactor Vessel Water Level Using Fuzzy Neural Networks in Severe Accidents due to LOCA

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soonho; Kim, Jaehawn; Na, Mangyun [Chosun Univ., Gwangju (Korea, Republic of)

    2013-05-15

    When the initial events that may lead to the severe accident such as Loss Of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) occurs at a nuclear power plant, it is most important to check the status of the plant conditions by observing the safety-related parameters such as neutron flux, pressurizer pressure, steam generator pressure and water level. In this paper, we propose a method of predicting the water level of coolant in the reactor vessel that directly affect the important events such as the exposure of the reactor core and the damage of reactor vessel by using a Fuzzy Neural Network (FNN) method. In addition, the data for verifying a proposed model was obtained by simulating the severe accident scenarios for the OPR1000 nuclear power plant using the MAAP4 code. In this paper, a prediction model was developed for predicting the reactor vessel water level using the FNN method. The proposed FNN model was verified based on the simulation data of OPR1000 by using MAAP4 code. As a result of simulation, we could see that the performance of the proposed FNN model is quite satisfactory but some large errors are observed occasionally. If the proposed FNN model is optimized by using a variety of data, it is possible to predict the reactor vessel water level exactly.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Institute of Scientific and Technical Information of China (English)

    陈长才; 宁可; 等

    1999-01-01

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

  18. Pilot program: NRC severe reactor accident incident response training manual: US Nuclear Regulatory Commission response

    International Nuclear Information System (INIS)

    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. US Nuclear Regulatory Commission Response is the fifth 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 NRC response modes, organizations, and official positions; roles of other federal agencies are also described briefly. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. 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. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-07-01

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

  1. Thermo-physical properties of corium: development of an assessed data base for severe accident applications

    Energy Technology Data Exchange (ETDEWEB)

    Strizhov, V.F.; Galimov, R.G.; Ozrin, V.D. [Nuclear Safety Institute of the Russian Academy of Sciences, Moscow (Russian Federation); Yu Zitserman, V.; Kobzev, G.I.; Fokin, L.R. [Institute of high temperatures, Russian Academy of Sciences, Moscow (Russian Federation); Piluso, P. [CEA Cadarache (DEN/DTN/STRI), Lab. d' essais pour la Maitrise des Accidents graves, 13 - Saint Paul lez Durance (France); Chalaye, H. [CEA Saclay, Dir. de l' Energie Nucleaire, 91 - Gif sur Yvette (France)

    2007-07-01

    In a hypothetical case of a core melt-down scenarios a very high temperature would be reached (up to 3000 K). In this case, the materials of the core and structural materials (fuel, cladding, metallic alloys, concrete, etc.) could melt to form complex and aggressive mixtures called corium. Modelling of severe accident phenomena, code development and assessments of nuclear safety require a reliable knowledge of the thermophysical properties of corium at wide temperature range (below solidus temperature, between solidus and liquidus temperature and above the liquidus temperature). Common Russian-French project ISTC 3078, has been devoted to the development, assessment and recommendation for the establishment of a reliable thermophysical data base for severe accident applications. The project consists of two tasks related to properties of pure metallic (U, Zr, Fe, Cr, Ni) and oxide (UO{sub 2}, U{sub 3}O{sub 8}, U{sub 4}O{sub 9}, NiO, ZrO{sub 2}, Cr{sub 2}O{sub 3}, FeO, Fe{sub 2}O{sub 3}, Fe{sub 3}O{sub 4}, Al{sub 2}O{sub 3}, CaO, MgO, SiO{sub 2}, HfO{sub 2}, CeO{sub 2}) components, and mixtures relevant to severe accident conditions. Three categories of data (on UPAK classification) were considered: experimental data, critically evaluated data, and predicted data. The data of the first category is a result of specific experiment, data of the second category is a result of the analysis of data consistency and co-processing (expert and statistical) obtained in several experiments, data of the third category are based on model estimates, using correlations between different physical properties. The process of assessing, review and development of recommendation is described in the paper and illustrated by examples on thermophysical properties. (authors)

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

    Directory of Open Access Journals (Sweden)

    Shi Xing-Wei

    2013-07-01

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

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

    Directory of Open Access Journals (Sweden)

    J. Arndt

    2012-01-01

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

  4. Development of a dose assessment computer code for the NPP severe accident

    International Nuclear Information System (INIS)

    A real-time emergency dose assessment computer code called KEDA (KAIST NPP Emergency Dose Assessment) has been developed for the NPP severe accident. A new mathematical model which can calculate cloud shine has been developed and implemented in the code. KEDA considers the specific Korean situations(complex topography, orientals' thyroid metabolism, continuous washout, etc.), and provides functions of dose-monitoring and automatic decision-making. To verify the code results, KEDA has been compared with an NRC officially certified code, RASCAL, for eight hypertical accident scenarios. Through the comparison, KEDA has been proved to provide reasonable results. Qualitative sensitivity analysis also the been performed for potentially important six input parameters, and the trends of the dose v.s. down-wind distance curve have been analyzed comparing with the physical phenomena occurred in the real atmosphere. The source term and meteorological conditions are turned out to be the most important input parameters. KEDA also has been applied to simulate Kori site and a hyperthetical accident with semi-real meteorological data has been simulated and analyzed

  5. 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)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  7. Mitigation of sodium-cooled fast reactor severe accident consequences using inherent safety principles

    International Nuclear Information System (INIS)

    Full text: 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. In the United States, accidents which have the potential for severe consequences usually are of probability less than 1 x 10-4 per reactor year, intended to satisfy the U.S. Nuclear Regulatory Commission (NRC) goal of limiting accidents with any fuel melting to such low probabilities. Such severe accidents include the category of Anticipated Transient Without Scram (ATWS) events mentioned above. Three accidents are usually analyzed to evaluate the reactor response in these cases; the unprotected (unscrammed) loss-of-flow (ULOF), where pumping power is lost and the pumps coast down, reducing coolant flow through the reactor core; the unprotected transient overpower (UTOP), where a control rod is inadvertently withdrawn from the core; and the unprotected loss-of-heat-sink (ULOHS), where the steam generator is isolated from the reactor in response to a turbine trip. For each of these 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

  8. A generic approach for steel containment vessel success criteria for severe accident loads

    International Nuclear Information System (INIS)

    Safety has been defined as the foremost design criterion for the Heavy Water New Production Reactor (NPR-HWR) by the U.S. DOE, Office of New Production Reactors (NP). The DOE-NP issued the Deterministic Severe Accident Criteria (DSAC) concept to guide the design of the NPR-HWR containment for resistance to severe accidents. The DSAC concept provides for a generic approach for containment vessel success criteria to predict the threshold of containment failure under severe accident loads. This concept consists of two parts: (1) Problem Statements and (2) Success Criteria. The paper is limited to a discussion of a success criteria. These criteria define acceptable containment response measures and limits for each problem statement. The criteria are based on the 'best estimate' of failure with no conservatism. Rather, conservatism, if required, is to be provided in the problem statements prepared by the designer and/or the regulatory authorities. The success criteria are presented on a multi-tiered basis for static pressure and temperature loadings, dynamic loadings, and missiles that may impact the containment. Within the static pressure and temperature loadings and the dynamic loadings, the criteria are separated into elastic analysis success criteria and inelastic analysis success criteria. Each of these areas, in turn, defines limits on either the stress or strain measures as well as on measures for buckling and displacements. The rationale upon which these criteria are based is contained in referenced documents. Rigorous validation of the criteria by comparison with results from analytical or experimental programs and application of the criteria to a containment design remain as future tasks. (orig./HP)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  10. Sustainable integration of EU research in severe accident phenomenology and management (SARNET2 project)

    International Nuclear Information System (INIS)

    In order to optimise the use of the available means and to constitute sustainable research groups in the European Union, the Severe Accident Research NETwork of Excellence (SARNET) has gathered 51 organisations representing most of the actors involved in Severe Accident (SA) research in Europe plus Canada. This project was co-funded by the European Commission (EC) under the 6th Euratom Framework Programme. Its objective was to resolve the most important pending issues for enhancing, in regard of SA, the safety of existing and future Nuclear Power Plants (NPPs). SARNET tackled the fragmentation that existed between the national R and D programmes, in defining common research programmes and developing common computer codes for safety assessment. The Joint Programme of Activities consisted in: (i) Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; (ii) Harmonizing and re-orienting the research programmes, and defining new ones; (iii) Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; (iv) Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by integrating the knowledge produced within SARNET; (v) Developing Scientific Databases, in which the results of research experimental programmes are stored in a common format; (vi) Developing a common methodology for Probabilistic Safety Assessment of NPPs; (vii) Developing short courses and writing a text book on Severe Accidents for students and researchers; (viii) Promoting personnel mobility amongst various European organizations. This paper presents the major achievements after four and a half years of operation of the network, in terms of knowledge gained, of improvements of the ASTEC reference code, of dissemination of results and of integration of the research programmes conducted by the various

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-10-15

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

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

    International Nuclear Information System (INIS)

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

  13. The importance of human performance and procedures in limiting severe accident risks

    International Nuclear Information System (INIS)

    Due to the defense in depth concept and redundancy in safety systems utilized, complex industrial plants, such as nuclear power plants (NPPs) can be operated safely. This capability has been demonstrated by many years of safe operation by numerous NPPs in the US and abroad. However, the occurrence of severe accidents has also demonstrated that constant vigilance in a number of areas is necessary to ensure continued safe operation. The areas noted as particularly important are Design, Organization and Management, Maintenance, and Operations (Human Performance). 18 refs

  14. [Fatal hyperpyrexia in an adolescent patient with severe burns after a traffic accident].

    Science.gov (United States)

    Jaehn, T; Sievers, R; Junger, A; Graunke, F; Blings, A; Reichert, B

    2016-07-01

    After a motorcycle accident a 16-year-old patient suffered severe burns to 40.5 % of the total body surface area (TBSA) of which 37 % were deep subdermal burns. After tangential and partly epifascial necrosectomy, Integra® was used as a temporary dermis replacement material for the lower extremities, combined with extensive negative pressure wound therapy (NPWT). In the further course of the treatment the patient developed uncontrollable hyperpyrexia with a fatal outcome. Possible influencing factors, such as the dermis replacement material combined with NPWT over large areas as well as the differential diagnoses propofol infusion syndrome, heatstroke and malignant hyperthermia are discussed. PMID:26767381

  15. Computational Aid of CANDU Reactor Severe Accident Management Strategies for Containment Integrity

    Energy Technology Data Exchange (ETDEWEB)

    Choi, Young; Kim, D. H.; Kim, S. D.; Park, S. Y.; Jin, Y. H. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Six computational aids (CA) have developed for the severe accident management strategies, plant specific features and behaviors must be studied by detailed analysis works. These computational aids (CA) are one of the tools that can be used to assess the hydrogen generation and challenges that may occur. This paper shows hydrogen related CA. The purpose of this CA is to define whether the hydrogen in the reactor building atmosphere is flammable, and to estimate the hydrogen concentration in the reactor building atmosphere based on an estimated oxidation percentage

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

    Institute of Scientific and Technical Information of China (English)

    XIAO Jianjun; ZHOU Zhiwei; JING Xingqing

    2006-01-01

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

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

    Science.gov (United States)

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

    2016-04-01

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

  18. Development of Lower Plenum Molten Pool Module of Severe Accident Analysis Code in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Shim, Suk-Ku; Marigomen, Ralph [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2014-10-15

    To simulate a severe accident progression of nuclear power plant and forecast reactor pressure vessel failure, we develop computational software called COMPASS (COre Meltdown Progression Accident Simulation Software) for whole physical phenomena inside the reactor pressure vessel from a core heat-up to a vessel failure. As a part of COMPASS project, in the first phase of COMPASS development (2011 - 2014), we focused on the molten pool behavior in the lower plenum, heat-up and ablation of reactor vessel wall. Input from the core module of COMPASS is relocated melt composition and mass in time. Molten pool behavior is described based on the lumped parameter model. Heat transfers in between oxidic, metallic molten pools, overlying water, steam and debris bed are considered in the present study. The models and correlations used in this study are appropriately selected by the physical conditions of severe accident progression. Interaction between molten pools and reactor vessel wall is also simulated based on the lumped parameter model. Heat transfers between oxidic pool, thin crust of oxidic pool and reactor vessel wall are considered and we solve simple energy balance equations for the crust thickness of oxidic pool and reactor vessel wall. As a result, we simulate a benchmark calculation for APR1400 nuclear power plant, with assumption of relocated mass from the core is constant in time such that 0.2ton/sec. We discuss about the molten pool behavior and wall ablation, to validate our models and correlations used in the COMPASS. Stand-alone SIMPLE program is developed as the lower plenum molten pool module for the COMPASS in-vessel severe accident analysis code. SIMPLE program formulates the mass and energy balance for water, steam, particulate debris bed, molten corium pools and oxidic crust from the first principle and uses models and correlations as the constitutive relations for the governing equations. Limited steam table and the material properties are provided

  19. Study of Containment Vent Strategies During Severe Accident Progression for the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout. Severe core damage occurred and a large amount of radioactive materials are released outside of the plants. After this terrible accident Nuclear Safety and Security Commission (NSSC) enforced to increase nuclear safety for all operating plants in Korea. To increase plant safety, both hardware reinforcement and software improvement are encouraged. Hardware reinforcement includes the preparation of the external water injection paths to the RCS and the spent fuel pool, a filtered containment venting system (CFVS), and AC power generating truck. Software improvement includes the increase of the effectiveness of the severe accident management guidance (SAMG) and plant staff training. To comply with NSSC's request, Wolsong Unit 1 has fulfilled the hardware reinforcement including the installation of a CFVS and started the extension of a SAMG to the low power and shutdown operation mode. Current SAMG deals accident occurred during full power operation only. The CFVS is designed to open and to close isolation valves manually. It does not require AC power. The operation of the CFVS prevents the reactor containment building failure due to the over-pressurization but it may release radioactive materials out of the reactor containment building. This paper discusses the radiological source terms for the containment vent strategy during severe accident progression which occurred during shutdown operation mode. This work is a part of the development of shutdown SAMG.. The CFVS is an effective means to control the containment pressure when the local air coolers are unavailable. Radioactive materials may release through the CFVS, but their amounts are reduced significantly. The alternative means, i.e., containment vent through the ventilation system which does not have an effective filter, is not a good choice to control the containment condition. It can maintain the containment

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

    Directory of Open Access Journals (Sweden)

    Sandro Paci

    2012-01-01

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

  1. Containment performance of prototypical reactor containments subjected to severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Klamerus, E.W.; Bohn, M.P. [Sandia National Labs., Albuquerque, NM (United States); Wesley, D.A. [EQE International, Irvine, CA (United States); Krishnaswamy, C.N. [Sargent & Lundy, Chicago, IL (United States)

    1996-12-01

    In SECY-90-016, the NTRC proposed a safety goal of a conditional containment failure probability (CCFP) of 0.1 and the alternative acceptance criteria allowed for steel containments, which specifies that the stresses should not exceed ASNE Level C allowables for severe accident pressures and temperatures. In this work, the need for an equivalent criterion for concrete containments was studied. Six surrogate containments were designed and analyzed in order to compare the margins between design pressure, pressure resulting in exceedance of Level C (or yield) stress limits, and ultimate pressure. For comparability, each containment has an identical internal volume and design pressure. Results from the analysis showed margins to yield are comparable and display a similar margin for both steel and concrete containments. In addition, the margin to failure, although slightly higher in the steel containments, were also comparable. Finally, a CCFP for code design was determined based on general membrane behavior and imposing an upper bound severe accident curve developed in the DCH studies. The resulting CCFP`s were less then 0.02 (or 2%) for all the surrogate containments studied, showing that these containment designs all achieved the NRC safety goal.

  2. Special committee review of the Nuclear Regulatory Commission's severe accident risks report (NUREG--1150)

    International Nuclear Information System (INIS)

    In April 1989, the Nuclear Regulatory Commission's (NRC) Office of Nuclear Regulatory Research (RES) published a draft report ''Severe Accident Risks: An Assessment for Five US Nuclear Power Plants,'' NUREG-1150. This report updated, extended and improved upon the information presented in the 1974 ''Reactor Safety Study,'' WASH-1400. Because the information in NUREG-1150 will play a significant role in implementing the NRC's Severe Accident Policy, its quality and credibility are of critical importance. Accordingly, the Commission requested that the RES conduct a peer review of NUREG-1150 to ensure that the methods, safety insights and conclusions presented are appropriate and adequately reflect the current state of knowledge with respect to reactor safety. To this end, RES formed a special committee in June of 1989 under the provisions of the Federal Advisory Committee Act. The Committee, composed of a group of recognized national and international experts in nuclear reactor safety, was charged with preparing a report reflecting their review of NUREG-1150 with respect to the adequacy of the methods, data, analysis and conclusions it set forth. The report which precedes reflects the results of this peer review

  3. Experimental study on hydrogen behavior in containment buildings filled with steam during severe accidents

    International Nuclear Information System (INIS)

    A severe nuclear power plant accident can result in the generation of large quantities of hydrogen that are then inevitably released into the containment building of the plant. As observed at the Fukushima Daiichi plant, hydrogen generated because of core degradation, when mixed with air, can lead to a hydrogen explosion and destroy the reactor building. Many studies have aimed to estimate the hydrogen distribution in environments both with and without steam. In this study, we investigate the hydrogen transport behavior in steam that condenses on the walls of the containment building, because this scenario is more representative of the actual conditions that would exist in the building after a severe accident. In our experiments, we used helium instead of hydrogen because it is stable and has a similar molar weight to hydrogen. After filling a cylinder, representative of the containment building, with steam, we found that steam condensed into water droplets because of heat transfer to the stainless steel wall and then ran down to the cylinder bottom. Helium sank to the vessel bottom by a condensation sedimentation effect. This transport phenomenon differed from that when there was no condensation. Helium sensors were used to monitor the transient helium concentration. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-09-15

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

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

    Directory of Open Access Journals (Sweden)

    Jean-Pierre Van Dorsselaere

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

  10. Overview of plant specific severe accident management strategies for Kozloduy nuclear power plant, WWER-1000/320

    Energy Technology Data Exchange (ETDEWEB)

    Andreeva, M. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: m_andreeva@inrne.bas.bg; Pavlova, M.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: pavlova@inrne.bas.bg; Groudev, P.P. [Institute for Nuclear Research and Nuclear Energy, Tzarigradsko Shaussee 72, Sofia 1784 (Bulgaria)], E-mail: pavlinpg@inrne.bas.bg

    2008-04-15

    This paper focuses on the fourth level of the defence in depth concept in nuclear safety, including the transitions from the third level and into the fifth level. The use of the severe accident management guideline (SAMG) is required when an accident situation is not handled adequately through the use of emergency operating procedures (EOP), thus leading to a partial or a total core melt. In the EOPs, the priority is to save the fuel, whereas, in the SAMG, the priority is to save the containment. Actions recommended in the SAMG aim at limiting the risk of radiologically significant radioactive releases in the short- and mid-term (a few hours to a few days). The paper describes basic severe accident management requirements related to nuclear power plant (NPP), specified by the IAEA and in Republic of Bulgaria Nuclear Legislation. It also surveys plant specific severe accident management (SAM) strategies for the Kozloduy NPP, equipped with WWER-1000 type reactors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

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

  12. THAI experiments on hydrogen and fission product behavior in the LWR containment during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Gupta, Sanjeev [Becker Technologies GmbH, Berlin (Germany)

    2012-03-15

    In case of a severe accident in a nuclear reactor, a large amount of hydrogen and fission products may be generated by interactions of the core melt with containment structures and water. The behaviour and the distribution of hydrogen and the fission products can be affected by a variety of chemical and physical phenomena taking place inside the containment. These phenomena involve the disciplines of thermal hydraulics, hydrogen distribution and deflagrations, fission products chemistry and material interactions, aerosol physics, and effectiveness of mitigation measures among others. Predictions of the consequences of a severe accident by conventional Lumped Parameter (LP) and Computational Fluid Dynamics (CFD) codes need to be based on large scale coupled-phenomenon experiments to minimize the scale effect in extrapolation to reactor safety analysis purpose. In this context, an extensive experimental program on nuclear severe accident has been pursuing at the THAI (Thermal-hydraulics, Hydrogen, Aerosol, and Iodine) test facility for many years. Main component of the facility is a 60-m{sup 3} stainless steel vessel, 9.2m high and 3.2m in diameter, with exchangeable internals for multi-compartment investigations. The test facility is operated by Becker Technologies under sponsorship of the German Federal Ministry of Economics and Technology. Since its construction in 2000, sponsorship of the German Federal Ministry of Economics and Technology. Since its construction in 2000, THAI facility has been engaged in the field of reactor safety in the frame of various national (THAI-I to THAI-IV) and international programs (OECD-THAI and THAI2 projects). Additionally, experimental data has been provided in the frame of several International Standard Problems (ISP 41, 46, 47, and 49) for code validation exercises. The THAI test facility allows investigating various accident scenarios, ranging from turbulent free convection to stagnant stratified containment atmospheres, and

  13. Experimental analyses of iodine behavior under severe accident conditions with ART

    Energy Technology Data Exchange (ETDEWEB)

    Hidaka, A.; Sugimoto, J. [Japan Atomic Energy Res. Inst., Ibaraki (Japan); Igarashi, M. [Kawasaki Heavy Industries, Ltd., 4-25, 2-Chome, Minami-suna, Koto-ku, Tokyo, 136 (Japan); Hashimoto, K. [Nuclear Power Engineering Corporation, 17-1, 3-Chome, Toranomon, Minato-ku, Tokyo, 105 (Japan); Yoshino, T. [Toshiba Advanced System Corporation, 1-2-4 Isago Kawasaki-ku, Kawasaki-shi, Kanagawa-ken, 210 (Japan)

    1997-09-01

    The fission products (FP) behavior analysis code ART developed at JAERI simulates a FP transport and deposition in a reactor coolant system and containment during severe accidents. As part of the code development and verification, several experimental analyses have been conducted. In the JAERI`s WAVE experiment, the effect of nitrogen or steam carrier gases on the cesium iodide (CsI) behavior in piping was recently investigated. The ART analysis for nitrogen agreed with the experimental results by reflecting the detailed thermo-fluiddynamic calculation on the CsI aerosol behavior analysis. On the contrary, the analysis for steam did not agree well with the experimental results because observed enhancement of aerosol growth cannot be explained by existing models. Moreover, the newly developed empirical models on iodine chemistry in water were examined for the ACE/RTF 3B experiment. The analysis showed that those models have a fundamental analytical capability. (orig.).

  14. Some aspects of the research and development programmes on the behaviour of containments during severe accidents

    International Nuclear Information System (INIS)

    The R and D programmes relating to the behaviour of containments during severe accidents cover several domains: .leaktightness of the containment: this programme concerns the mechanical resistance of the concretes and the cracking criteria, on the one hand, and the leak rate through the porosities or cracks, on the other; . gaseous releases inside the containment. In addition to the releases of steam and fission products from the primary circuit, the gaseous H20 and C02 releases from the concrete must also be studied: firstly during the corium-concrete interaction, and secondly during the heating of the internal surface of the containment which can be raised to a high temperature on contact with the atmosphere, for example during hydrogen combustion; . the release of fission products during the corium-concrete interactions; . the behaviour of the fission products inside the containment, particularly as regards iodine

  15. Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

    International Nuclear Information System (INIS)

    Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

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

    International Nuclear Information System (INIS)

    This document reviews the current state of research on severe accidents in France and other countries. It aims to provide an objective vision, and one that's as exhaustive as possible, for this innovative field of research. It will help in identifying R and D requirements and categorising them hierarchically. Obviously, the resulting prioritisation must be completed by a rigorous examination of needs in terms of safety analyses for various risks and physical phenomena, especially in relation to Level 2 Probabilistic Safety Assessments. PSA-2 should be sufficiently advanced so as not to obscure physical phenomena that, if not properly understood, might result in substantial uncertainty. It should be noted that neither the safety analyses nor PSA-2 are presented in this document. This report describes the physical phenomena liable to occur during a severe accident, in the reactor vessel and the containment. It presents accident sequences and methods for limiting impact. The corresponding scenarios are detailed in Chapter 2. Chapter 3 deals with in-vessel accident progression, examining core degradation (3.1), corium behaviour in the lower head (3.2), vessel rupture (3.3) and high-pressure core meltdown (3.4). Chapter 4 focuses on phenomena liable to induce early containment failure, namely direct containment heating (4.1), hydrogen risk (4.2) and steam explosions (4.3). The phenomenon that could lead to a late containment failure, namely molten core-concrete interaction, is discussed in Chapter 5. Chapter 6 focuses on problems related to in-vessel and ex-vessel corium retention and cooling, namely in-vessel retention by flooding the primary circuit or the reactor pit (6.1), cooling of the corium under water during the corium-concrete interaction (6.2), corium spreading (6.3) and ex-vessel core catchers (6.4). Chapter 7 relates to the release and transport of fission products (FP), addressing the themes of in-vessel FP release (7.1) and ex-vessel FP release (7.3), FP

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

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

  18. A Novel Thermal-Mechanical Detection System for Reactor Pressure Vessel Bottom Failure Monitoring in Severe Accidents

    International Nuclear Information System (INIS)

    Following the Fukushima Daiichi nuclear accident in Japan, there is an increased need of enhanced capabilities for severe accident management (SAM) program. Among others, a reliable method for detecting reactor pressure vessel (RPV) bottom failure has been evaluated as imperative by many utility owners. Though radiation and/or temperature measurement are potential solutions by tradition, there are some limitations for them to function desirably in such severe accident as that in Japan. To provide reliable information for assessment of accident progress in SAM program, in this paper we propose a novel thermal-mechanical detection system (TMDS) for RPV bottom failure monitoring in severe accidents. The main components of TMDS include thermally sensitive element, metallic cables, tension controlled switch and main control room annunciation device. With TMDS installed, there shall be a reliable means of keeping SAM decision-makers informed whether the RPV bottom has indeed failed. Such assurance definitely guarantees enhancement of severe accident management performance and significantly improve nuclear safety and thus protect the society and people. (authors)

  19. An overview of severe accident modeling and analysis work for the ANS reactor conceptual safety analysis report

    International Nuclear Information System (INIS)

    ORNL's Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management has initiated severe accident analysis and related technology development efforts early-on in the design phase itself. Early consideration of severe accident issues will aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It will also provide a means for satisfying on- and off-site regulatory requirements and provide containment response and source term analyses for level-2 and -3 Probabilistic Risk Analyses (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions, and consequently provide insights for the development of strategies and design philosophies for accident management, mitigation, and emergency preparedness. This paper presents a perspective overview of the severe accident modeling and analysis work for the ANS Conceptual Safety Analysis Report (CSAR)

  20. Severe accidents in spent fuel pools in support of generic safety, Issue 82

    International Nuclear Information System (INIS)

    This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified

  1. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    Energy Technology Data Exchange (ETDEWEB)

    Audin, L.

    1990-12-01

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs.

  2. The study of steam explosions in nuclear systems. Advanced Reactor Severe Accident Program

    International Nuclear Information System (INIS)

    This report presents an overview of the steam explosion issue in nuclear reactor safety and our approach to assessing it. Key physics, models, and computational tools are described, and illustrative results are presented for ex-vessel steam explosions in an open pool geometry. An extensive set of appendices facilitate access to previously reported work that is an integral part of this effort. These appendices include key developments in our approach, key advances in our understanding from physical and numerical experiments, and details of the most advanced computational results presented in this report. Of major significance are the following features: A consistent two-dimensional treatment for both premixing and propagation which in practical settings are ostensibly at least two-dimensional phenomena; experimental demonstration of voiding and microinteractions which represent key behaviors in premixing and propagation respectively; demonstration of the explosion venting phenomena in open pool geometries which, therefore, can be counted on as a very important mitigative feature; and introduction of the idea of penetration cutoff as a key mechanism prohibiting large-scale premixing in usual ex-vessel situations involving high pour velocities and subcooled pools. This report is intended as an overview and is to be followed by code manuals for PM-ALPHA and ESPROSE.m, respective verification reports, and application documents for reactor-specific applications. The applications will employ the Risk Oriented Accident Analysis Methodology (ROAAM) to address the safety importance of potential steam explosions phenomena in evaluated severe accidents for passive Advanced Light Water Reactors (ALWRs)

  3. Response of a LWR pressure vessel to severe-accident loadings

    Energy Technology Data Exchange (ETDEWEB)

    Ju, F.D.; Bennett, J.G.; Anderson, C.A.

    1982-01-01

    In the recent emphasis on nuclear safety, structural studies of nuclear reactor vessels have been directed toward evaluating their response during severe loading incidents or accidents including even core meltdown - however improbable these accidents may be. The present paper will address some of these problems. The ultimate load carrying capacity of an unflawed nuclear pressure vessel is estimated. The measure of the maximum pressure that the vessel can resist during quasistatic loading is a useful quantitative estimate of overall vessel strength. The paper than analyzes two structural problems during a hypothetical meltdown. In the initial stage, the molten core mixture drops into the lower portion of the pressure vessel, resulting in both temperature and pressure rises. Subsequently, a vapor explosion may occur as a result of the molten metal coming in sudden contact with the water in the lower portion of the vessel. The explosion is postulated to propel a slug of molten metalup the vessel barrel that eventually impacts the upper head of the vessel potentially generating missiles in the containment building. The reactor vessel at Indian Point, New York is used as a prototype of this analysis.

  4. Response of a LWR pressure vessel to severe-accident loadings

    International Nuclear Information System (INIS)

    In the recent emphasis on nuclear safety, structural studies of nuclear reactor vessels have been directed toward evaluating their response during severe loading incidents or accidents including even core meltdown - however improbable these accidents may be. The present paper will address some of these problems. The ultimate load carrying capacity of an unflawed nuclear pressure vessel is estimated. The measure of the maximum pressure that the vessel can resist during quasistatic loading is a useful quantitative estimate of overall vessel strength. The paper than analyzes two structural problems during a hypothetical meltdown. In the initial stage, the molten core mixture drops into the lower portion of the pressure vessel, resulting in both temperature and pressure rises. Subsequently, a vapor explosion may occur as a result of the molten metal coming in sudden contact with the water in the lower portion of the vessel. The explosion is postulated to propel a slug of molten metalup the vessel barrel that eventually impacts the upper head of the vessel potentially generating missiles in the containment building. The reactor vessel at Indian Point, New York is used as a prototype of this analysis

  5. 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.

  6. Severe accidents in spent fuel pools in support of generic safety, Issue 82

    Energy Technology Data Exchange (ETDEWEB)

    Sailor, V.L.; Perkins, K.R.; Weeks, J.R.; Connell, H.R.

    1987-07-01

    This investigation provides an assessment of the likelihood and consequences of a severe accident in a spent fuel storage pool - the complete draining of the pool. Potential mechanisms and conditions for failure of the spent fuel, and the subsequent release of the fission products, are identified. Two older PWR and BWR spent fuel storage pool designs are considered based on a preliminary screening study which tried to identify vulnerabilities. Internal and external events and accidents are assessed. Conditions which could lead to failure of the spent fuel Zircaloy cladding as a result of cladding rupture or as a result of a self-sustaining oxidation reaction are presented. Propagation of a cladding fire to older stored fuel assemblies is evaluated. Spent fuel pool fission product inventory is estimated and the releases and consequences for the various cladding scenarios are provided. Possible preventive or mitigative measures are qualitatively evaluated. The uncertainties in the risk estimate are large, and areas where additional evaluations are needed to reduce uncertainty are identified.

  7. Analysis of steam generator tube rupture as a severe accident using MELCOR 1.8.4

    Energy Technology Data Exchange (ETDEWEB)

    Yang Hongrun; Hidaka, Akihide; Sugimoto, Jun [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

    1999-03-01

    This report presents the results from the MELCOR 1.8.4 calculations for Steam Generator Tube Rupture (SGTR) with stuck open of all the safety valves in faulted SG as a severe accident. The calculations are based on Surry nuclear power plant. After performed using the once-through primary system model alone by 1.0x10{sup 5} s, the calculations were conducted with both of the once-through and the hot leg countercurrent natural circulation models. The results, including event sequences, processes and progressions of core degradation, radionuclides release from core and reactor cavity, and source terms to the environment are described in detail. It is concluded that the availability of High Pressure Safety Injection (HPSI) can significantly delay the progression of core heat-up and approximately 7% of cesium iodide (CsI) can be released to the environment directly through the stuck open safety valve. Comparisons between the results from the two models are also given in this report. The present analyses also showed that during SGTR accident, the hot leg countercurrent natural circulation flow cannot be established well and therefore it has little effect on the mitigation of the core degradation. (author)

  8. Nuclear waste shipping container response to severe accident conditions, A brief critique of the modal study

    International Nuclear Information System (INIS)

    The Modal Study (NUREG/CR-4829) attempts to upgrade the analysis of spent nuclear fuel transportation accidents, and to verify the validity of the present regulatory scheme of cask performance standards as a means to minimize risk. While an improvement over many prior efforts in this area (such as NUREG-0170), it unfortunately fails to create a realistic simulation either of a shipping cask, the severe conditions to which it could be subjected, or the potential damage to the spent fuel cargo during an accident. There are too many deficiencies in its analysis to allow acceptance of its results for the presumed cask design, and many pending changes in new containers, cargoes and shipping patterns will limit applicability of the Modal Study to future shipments. In essence, the Modal Study is a good start, but is too simplistic, incomplete, outdated and open to serious question to be used as the basis for any present-day environmental or risk assessment of spent fuel transportation. It needs to be redone, with peer review during its production and experimental verification of its assumptions, before it has any relevance to the shipments planned to Yucca Mountain. Finally, it must be expanded into a full risk assessment by inputing its radiological release fractions and probabilities into a valid dispersal simulation to properly determine the impact of its results. 51 refs

  9. Pilot program: NRC [Nuclear Regulatory Commission] severe reactor accident incident response training manual: Overview and summary of major points

    International Nuclear Information System (INIS)

    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. Overview and Summary of Major Points is the first 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 assessment. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance of license requirements for NRC licensees. 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

  10. APRI-7 Accident Phenomena of Risk Importance. A progress report on research in the field of severe accidents in 2009-2011

    International Nuclear Information System (INIS)

    Knowledge of the phenomena that may occur during severe accidents in a nuclear power plant is an important prerequisite for being able to predict the plant behavior, in order to formulate procedures and instructions for incident handling, for contingency planning, and to get good quality at the accident analysis and risk studies. Since the early 80's nuclear power companies and authorities in Sweden has collaborated in research on severe reactor accidents. Cooperation in the beginning was mostly linked to strengthening the protection against environmental impacts after a severe reactor accident, in particular to develop systems for filtered depressurization of the reactor containment. Since the early 90's the cooperation has partially changed and shifted to the phenomenological questions of risk dominance. During the years 2009-2011, cooperation continued in the research-program APRI-7. The aim was to show whether the solutions adopted in the Swedish strategy for accident management provides reasonable protection for the environment. This was done by gaining detailed knowledge of both important phenomena in the hearth melting behavior, and the amount of radioactivity that can be discharged to the surroundings during a severe accident. To achieve this aim, the research program has included a follow-up of international research in severe accidents and evaluation of results, and continued to support research at KTH and Chalmers Univ. of severe accidents. The follow-up of international research has promoted the exchange of knowledge and experience and has provided access to a wealth of information about various phenomena relevant to the events at severe accidents. This was important to obtain a good basis for assessment of abatement measures in the Swedish nuclear reactors. Continuing support to the Royal Inst. of Technology has provided increased knowledge about the ability to cool the molten core of the reactor vessel and the processes associated with cooling the

  11. 'SAMIME' - An EC Concerted Action on Severe Accident Management in Europe

    International Nuclear Information System (INIS)

    The EC Concerted Action 'Severe Accident Management Implementation and Expertise' - with acronym 'SAMIME' - was initiated in 1998 with the following main objectives: 1. to determine the status and the extent of the severe accident management (SAM) development and implementation in partner countries/regions/utilities; 2. to determine the extent to which the development is institutional and develop a consensus opinion among partners as to which elements are needed or useful in this respect; 3. to review the tools which are available to support SAM development and implementation and determine the extent to which these tools have a value during a SAM event; 4. to determine in which areas further research may be beneficial for SAM development. Thirteen organisations from all countries with NPPs in the European Union and neighbouring countries, also in Eastern Europe, agreed to participate in this effort. Partners were a balanced group of utilities, vendors and regulators. Two organisations became partners later on. One utility could not be a formal partner, but supported the project with its insights. Two US Owners Groups and the OECD Halden Reactor Programme provided input on specific topics. The Concerted Action was executed in a number of workshops which were preceded by extensive questionnaires to facilitate discussions. The project provided a good overview of what SAM guidance (SAMG) was in place in the various countries and in which direction the development went. After extensive discussions, a consensus opinion was achieved on what partners felt the elements of an adequate SAMG approach should be. Although differences may exist in the way SAMG can be implemented, a common base line could be defined - which was more than a common denominator, as it sometimes exceeded what was in place. Finally, the areas were outlined where additional research work still could enhance the SAMG, taking note of the fact that understanding of severe accidents would never be complete

  12. 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

  13. PSA-based expert system as an operator assistance during severe accidents

    International Nuclear Information System (INIS)

    In this work, an Expert System is presented, designed to assist the operator of the CAREM-25 Power Station during postulated Severe Accidents. CAREM-25 is a new generation, very low power (100 MWth, 25 Mwe net) nuclear station. The knowledge base of the Expert System was constructed from the Containment Events Trees of the corresponding PSA made for this Power Station. At the same time, the calculations of probabilities were improved. They were implemented in the different branches of the Containment Events Trees using fuzzy arithmetic. The Expert System has been designed in Fuzzy CLIPS V6.04A. In its present state of development it interacts with the user by asking questions that help to determine the state of damage of the plant. Like answers, it gives the data that characterize the state of the plant and the probable states at which the containment would arrive as well as the values from the associated probability. (author)

  14. Investigation of alternative solutions for severe accident management in future reactors

    International Nuclear Information System (INIS)

    Since 1991, the CEA/DRN 'Innovations-Future Reactors' Program (IFRP) has been developed in order to elaborate, to evaluate and validate technical options which can be of interest for future reactors. The main objectives of this program are: to improve both the safety and cost of future nuclear power plants, to optimize the fuel cycle and the management of nuclear materials. The present paper is focused on the third R and D theme, i.e., on the 'Innovation-Severe Accident Research Program' (ISARP). This specific CEA long-term program is developed in addition to shorter-term studies conducted in collaboration with the CEA partners (EDF and FRAMATOME), more particularly, for the future European Pressurized Water Reactor (EPR). (J.P.N.)

  15. Silver-indium-cadmium control rod behaviour during a severe reactor accident

    International Nuclear Information System (INIS)

    An alloy of silver, indium and cadmium is commonly used as control rod material in pressurised water reactors (PWRs). The behaviour of this alloy has been studied in a series of experiments using an induction furnace to achieve temperatures up to 1900K. The aerosols released from overheated clad and unclad control rod samples have been characterised in both steam and inert atmospheres. Mass balance experiments have been undertaken to determine the distribution of the control rod alloy constituents following rupture of the cladding, and this work has been supported by thermogravimetric studies of silver-indium mixtures. Metallographic studies were also undertaken to assess the failure mode of the stainless steel cladding and the interaction of the molten alloy with Zircaloy. The results of this work are discussed in terms of aerosol/vapour behaviour during severe reactor accidents. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-10-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-06-15

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

  18. 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.

  19. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident

    International Nuclear Information System (INIS)

    The report describes the fundamentals of heat conductivity and convection and numerical methods like finite difference and control volume method for calculation of the thermal history of a reactor pressure vessel bottom head and penetrations. Phase changes from solids to liquids are considered. Time integration is performed by explicit or implicit method. Developed computer codes for thermal conductivity and convection analyses and codes for graphical visualization are described. The codes are applied to two practical cases. They deal with analyses of Swiss CORVIS-experiments and analyses of control rod and instrument penetrations in a BWR bottom head. A model for calculation of effective thermal conductivity of granular corium is developed. The work is also related to EU MVI-project (Core Melt-Pressure Vessel Interactions During a Light Water Reactor Severe Accident), whose coordinator is Prof. B. R. Sehgal at Royal Institute of Technology in Stockholm. (orig.) (11 refs.)

  20. Agricultural measures to reduce radiation doses to man caused by severe nuclear accidents

    International Nuclear Information System (INIS)

    Agricultural land and products may become contaminated after a severe nuclear accident. If radiation doses to man caused by the ingestion of contaminated agricultural products from such areas will be unacceptably high, measures to reduce this radiation dose will have to be taken. Radiation doses to man can be estimated by using models which describe quantitatively the transfer of radionuclides through the biosphere. The following processes and pathways are described in this study: accidental releases into atmospheric environments and subsequent nearby deposition; contamination of crops by direct deposition and the subsequent short term pathway (e.g. grass-cow-milk-man); contamination of soil and the subsequent long term pathway (e.g. soil-crop-man, soil-grass-cattle-milk/meat-man). Depending on the degree of contamination and on the estimated radiation doses to man, various measures are advised. (Auth.)

  1. Human factors issues in severe accident management: Training for decision-making under stress

    International Nuclear Information System (INIS)

    Training for operator and other technical positions in the commercial nuclear power industry traditionally has focused on mastery of the formal procedures used to control plant systems and processes. However, there is a growing awareness that the decision-making tasks required for selecting appropriate control actions, in addition to guidance from formal procedures, also involve cognitive activities commonly referred to as judgment or reasoning. A project was completed to address the nature of the cognitive skills that may be important to decision-making in the nuclear power plant environment, especially during severe accident management. The project identified a model of decision-making that could account for both rule-based and knowledge-based decision-making and used it to identify cognitive skills for both individuals and operational crews. This analysis was then used to identify existing training techniques for cognitive skills and the general characteristics of successful training techniques

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  4. Preservation of thermalhydraulic and severe accident experimental data produced by the European Commission

    Energy Technology Data Exchange (ETDEWEB)

    Pla, Patricia; Ammirabile, Luca; Pascal, Ghislain; Annunziato, Alessandro [European Commission Joint Research Centre, Petten (Netherlands). Inst. for Energy and Transport; European Commission Joint Research Centre, Ispra (Italy). Inst. for the Protection and Security of the Citizen

    2014-07-15

    The experimental data recorded in Integral Effect Test Facilities (ITFs) are traditionally used in order to validate Best Estimate (BE) system codes and to investigate the behaviour of Nuclear Power Plants (NPP) under accident scenarios. In the same way, facilities dedicated to specific thermalhydraulic (TH) Severe Accident (SA) phenomena are used for the development and improvement of specific analytical models and codes used in the SA analysis for Light Water Reactors (LWR). The extent to which the existing reactor safety experimental databases are preserved was well known and frequently debated and questioned in the nuclear community. The Joint Research Centre (JRC) of the European Commission (EC) has been deeply involved in several projects for experimental data production and experimental data preservation. The paper is presenting these large EC initiatives on the production of experimental data and its storage in the JRC STRESA node. The objective of the paper is to further disseminate and promote the usage of the database containing these experimental data and to demonstrate long-term importance of well maintained experimental databases. At present time the Nuclear Reactor Safety Assessment Unit (NRSA) of the JRC Institute of Energy and Transport in Petten is engaged in the development of a new STRESA tool to secure EU storage for SA experimental data and calculations. The target is to keep the main features of the existing STRESA structure but using the new informatics technologies that are nowadays available and providing new capabilities. The development of this new STRESA tool should be completed by the end of 2014. (orig.)

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

    Directory of Open Access Journals (Sweden)

    MohammadJavad Jafari

    2015-10-01

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

  6. Effectiveness of In-Vessel Retention Strategies and Minimum Safety Injection Flow over Postulated Severe Accidents of OPR1000

    International Nuclear Information System (INIS)

    The objective of this study is first to evaluate various serious severe accident scenarios of OPR1000 with and without in-vessel retention strategies using MELCOR code. Second is to develop a mechanistic model of minimum safety injection flow using the thermal-hydraulic parameters of CET and collapsed water level obtained from the MELCOR simulation results. Effectiveness of RCS depressurization of OPR1000 is investigated for postulated severe accidents of SBLOCA, SBO, and TLOF. It is seen that timely operator action is important to achieve the best mitigation. Also The MELCOR simulation results of SBLOCA, SBO, and TLOFW are utilized to develop a model for minimum safety injection flow. The model suggests that if HPSI is available with RCS pressure lower than 120 bars, the core coolability can be guaranteed. In this study, several MELCOR simulations are conducted in search for effective in-vessel retention strategies over postulated severe accidents of SBLOCA, SBO, and TLOFW of OPR1000. Detailed accident sequences are presented and indicative parameters diagnosing the reactor thermal-hydraulic state are interrogated to provide useful information to the operator actions. To properly assist operator's action during the severe accident, the thermal-hydraulic parameters should be virtual, intuitive, and reliable. In addition, the parameters should be collected through the instrumentations close to the reactor core. In this regard, Core Exit Temperature (CET) and collapsed core water level are deemed as the commensurate parameters

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  8. Precursors to potential severe core damage accidents: 1996. A status report. Volume 25

    International Nuclear Information System (INIS)

    This report describes the 14 operational events in 1996 that affected 13 commercial light-water reactors and that are considered to be precursors to potential severe core damage accidents. All these events had conditional probabilities of subsequent severe core damage greater than or equal to 1.0 x 10-6. These events were identified by first computer-screening the 1996 licensee event reports from commercial light-water reactors to identify those events that could potentially be precursors. Candidate precursors were selected and evaluated in a process similar to that used in previous assessments. Selected events underwent engineering evaluation that identified, analyzed, and documented the precursors. Other events designated by the Nuclear Regulatory Commission (NRC) also underwent a similar evaluation. Finally, documented precursors were submitted for review by licensees and NRC headquarters and regional offices to ensure the plant design and its response to the precursor were correctly characterized. This study is a continuation of earlier work, which evaluated 1969--1995 events. The report discusses the general rationale for this study, the selection and documentation of events as precursors, and the estimation of conditional probabilities of subsequent severe core damage for the events

  9. Analysis of severe accident progression for in-vessel corium retention estimation in the APR 1400

    International Nuclear Information System (INIS)

    The scope and content of this technical report is to evaluate high-pressure transients of total Loss of Feed Water (LOFW) to the steam generators and Station Blackout (SBO), and low-pressure transients of Loss of Coolant Accident (LOCA) without Safety Injection (SI) using the SCDAP/RELAP5/MOD3.3 computer code from transient initiation to reactor vessel failure in the APR 1400. The SCDAP/RELAP5/MOD3.3 results have shown that the pressurizer surge line had failed before reactor vessel failure, which results in a rapid decrease of RCS pressure in the high-pressure sequences of the LOFW and the SBO transients. The LOFW with intentional RCS depressurization using the safety depressurization system prevents failure of the pressurizer surge line and results in actuation of the safety injection tanks. A large mass of the melted and relocated core material in the bottom of core region at approximately 6 hours was relocated to the lower plenum of the reactor vessel in the 2-inch and the 3-inch SBLOCAs, which results in the reactor vessel failure by creep. In the SBLOCA sequence without the safety injection, the actuation of the SITs can be possible for the operator to have time of 4-5 hours in the action of the severe accident mitigation strategy to prevent reactor vessel failure. In all sequences, approximately 50-90 % of the core material was melted and relocated to the lower plenum of the reactor vessel at the time of reactor vessel failure and approximately 30-60 % of the fuel rod cladding was oxidized

  10. A methodology for supporting decisions on the establishment of protective measures after severe nuclear accidents

    International Nuclear Information System (INIS)

    The objective of this report is to demonstrate the use of a methology supporting decisions on protective measures following severe nuclear accidents. A multicriteria decision analysis approach is recommended where value tradeoffs are postponed until the very last stage of the decision process. Use of efficient frontiers is made to exclude all technically inferior solutions and present the decision maker with all nondominated solutions. A choice among these solutions implies a value trade-off among the multiple criteria. An interactive computer packge has been developed where the decision maker can choose a point on the efficient frontier in the consequence space and immediately see the alternative in the decision space resulting in the chosen consequences. The methodology is demonstrated through an application on the choice among possible protective measures in contaminated areas of the former USSR after the Chernobyl accident. Two distinct cases are considered: First a decision is to be made only on the basis of the level of soil contamination with Cs-137 and the total cost of the chosen protective policy; Next the decision is based on the geographic dimension of the contamination ant the total cost. Three alternative countermeasure actions are considered for population segments living on soil contaminated at a certain level or in a specific geographic region: (a) relocation of the population; (b) improvement of the living conditions; and, (c) no countermeasures at all. This is final deliverable of the CEC-CIS Joint Study Project 2, Task 5: Decision-Aiding-System for Establishing Intervention Levels, performed under Contracts COSU-CT91-0007 and COSU-CT92-0021 with the Commission of European Communities through CEPN

  11. The effect of airborne hygroscopic matter on aerosol behavior in severe nuclear power plant accidents

    International Nuclear Information System (INIS)

    The growth of hygroscopic particles due to steam condensation has been studied. Sophisticated and simplified models for particle growth have been compared to see what sort of approximations can be used in the safety analyses of severe LWR accidents. A model suitable for evaluating condensational growth in containment conditions has been developed and incorporated in the NAUA containment aerosol program. The hygroscopic growth has been studied experimentally and the results of the NAUA-HYGROS computer program developed here have been compared with the experimental results. For heat and mass transport the classical Mason equation compared well with the more accurate solutions at the expected thermal-hydraulic conditions in the containment atmosphere. For applications with short residence times at high humidities, where the Mason equation is not strictly valid, a more accurate solution for mass and heat transfer to the particle has been developed. A conservative model for the dissolution of multicomponent particles at high relative humidities has been developed and included in the NAUA-HYGROS code. The development of NAUA-HYGROS was closely related to the LACE experiments, where the behavior of hygroscopic CsOH and non-hygroscopic MnO aerosols in a 852 m3 containment vessel was studied. The code results agreed well with the measured values. The comparison of the results from the Modular Accident Analyses Program (MAAP) with the LACE experimental results showed the invalidity of the simple model used. An experimental set-up for measuring the behavior of hygroscopic particles in a flow type reaction chamber was developed and constructed. The time development of initially dry CsOH and NaOH particles was measured at different relative humidities. The most critical parameters affecting the particle size were found to be the relative humidity and the residence time in the chamber before sampling. Theoretical calculations with the NAUA-HYGROS code showed that during travel

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    Institute of Scientific and Technical Information of China (English)

    TAO Jun; CAO Xuewu

    2009-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Kambiz Masoumi

    2014-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  16. Procedural and submittal guidance for the individual plant examination of external events (IPEEE) for severe accident vulnerabilities

    International Nuclear Information System (INIS)

    Based on a Policy statement on Severe Accidents, the licensee of each nuclear power plant is requested to perform an individual plant examination. The plant examination systematically looks for vulnerabilities to severe accidents and cost-effective safety improvements that reduce or eliminate the important vulnerabilities. This document presents guidance for performing and reporting the results of the individual plant examination of external events (IPEEE). The guidance for reporting the results of the individual plant examination of internal events (IPE) is presented in NUREG-1335. 53 refs., 1 figs., 2 tabs

  17. Fission-product chemistry in severe reactor accidents: Review of relevant integral experiments

    International Nuclear Information System (INIS)

    The attenuation of the radioactive fission-product emission from a severe reactor accident will depend on a combination of chemical, physical and thermal-hydraulic effects. Chemical species stabilised under the prevailing conditions will determine the extent of aerosol formation and any subsequent interaction, so defining the magnitude and physical forms of the eventual release into the environment. While several important integral tests have taken place in recent years, these experiments have tended to focus on the generation of mass-balance and aerosol-related data to test and validate materials-transport codes rather than study the impact of important chemical phenomena. This emphasis on thermal hydraulics, fuel behaviour and aerosol properties has occurred in many test (e.g. PBF, DEMONA, Marviken-V, LACE and ACE). Nevertheless, the generation and reaction of the chemical species in all of these programmes determined the transport properties of the resulting vapours and aerosols. Chemical effects have been studied in measurements somewhat subsidiary to the main aims of the tests. This work has been reviewed in detail with respect to Marviken-V, LACE, ACE and Falcon. Specific issues remain to be addressed, and these are discussed in terms of the proposed Phebus-FB programme. (author). 58 refs, 9 figs, 1 tab

  18. Applicability of Phebus FP results to severe accident safety evaluations and management measures

    International Nuclear Information System (INIS)

    The international Phebus FP (Fission Product) programme is the largest research programme in the world investigating core degradation and radioactive product release should a core meltdown accident occur in a light water reactor plant. Three integral experiments have already been performed. The experimental database obtained so far contains a wealth of information to validate the computer codes used for safety and accident management assessment

  19. French regulatory requirements for the occupational radiation protection in severe accident situations and post-accident recovery

    International Nuclear Information System (INIS)

    Workers of the concerned company and other persons and teams called 'intervention personnel' (specialized firemen, first aider, etc.) are to be involved in radiological emergency situations. Radiation protection provisions for workers and for intervention personnel complement one another because they cover persons with different statutes (workers under the responsibility of an employer and persons acting within the framework of agreements with the public authorities or within the framework of the requisitions). Work or operations exposing workers to ionizing radiation in radiological emergency situations can be assigned only to workers satisfying all of the following conditions: classification in category A worker; free of any medical unfitness; on a list drawn up in advance for this purpose; having received appropriate information on the risks and the precautions to take during the work or the operation; not having received, during the preceding twelve months, a dose greater than one of the annual limit values for exposures subject to special authorization. In addition, the worker must be a volunteer to carry out the work or the operations concerned in radiological emergency situations and have individual dosimetry means appropriate for the situation. Intervention personnel are possibly composed of personnel from responding organizations, such as police officers, fire-fighters, medical personnel, drivers and crews of evacuation vehicles, or of workers employed by the head of the damaged plant. In order to determine their selection, training and medical and radiological monitoring conditions, intervention personnel are classified into two groups: personnel forming the special technical, medical and health intervention teams readied in advance to deal with radiological emergency situations and persons not belonging to special teams but intervening as part of the tasks within the scope of their competence. In case of an existing exposure situation (post-accident

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-01-15

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

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

    Science.gov (United States)

    Pinnington, Sarah; Atterton, Brigid; Ingleby, Sarah

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  3. Study of top reflooding in case of severe accident and in particular oxidation of Uranium, Zirconium, Oxygen melts

    International Nuclear Information System (INIS)

    In 1979, the Three Mile Island (TMI) accident occurred in United States and accelerated research activities in the field of severe accidents. Severe accident management procedures imply massive water injections to flood the core. The work of this thesis bent principally over this reflooding. The first part of the study concerns the core oxidation enhancement during the reflooding phase which leads to a rough increase of the concentration of burnable hydrogen in the containment. This is why the study carried on the analysis of the contribution of the oxidation of U-Zr-O mixtures, towards the total production of hydrogen during reflooding. In the second part, the study concerns top flooding modelling i.e.: with injection of water in the hot legs. Here, we attempted to define bases and realize a model allowing to describe this type of reflooding. These models were validated on the simulation of the parameter with MAAP4 code. (author)

  4. Hydrogen management and the metamorphosis of NRC policy on severe nuclear accident risk

    International Nuclear Information System (INIS)

    From the early days of light water reactor developments, it was understood that, following a loss-of-coolant accident, hydrogen could accumulate inside the primary reactor containment as a result of: (1) metal-water reaction involving the fuel element cladding; (2) the radiolytic decomposition of the water in the reactor core and the containment sump; (3) the corrosion of certain construction materials by some spray solutions; and (4) possible synergistic effects of chemical, thermal and radiolytic by-products of accidents on containment protective coatings and electric cable insulation. The NRC's policy decisions regarding hydrogen management prior to and in light of the TMI-2 loss of coolant accident are discussed

  5. Thermal-hydraulic analysis on Ex-Vessel fuel Storage Tank of MONJU at severe accident

    International Nuclear Information System (INIS)

    In this paper, results of a thermal-hydraulic analysis on the Ex-Vessel fuel Storage Tank (EVST) of the fast breeder reactor MONJU at severe accident is described. Safety evaluations on this facility have ever been performed by using a one-dimensional flow network code. However, validation on a model of this code has been needed, because EVST has plenums and asymmetry equipment. Therefore we performed a CFD analysis under a condition of station blackout (SBO) in order to clarify the circulation flow rate and multidimensionality of the EVST. As a result, the following points were confirmed: 1) Circulation flow rate is maintained half of a flow rate at the rated operation condition at the minimum. 2) Thermal stratification arises in the lower plenum at SBO. 3) Circumferential distribution of flow rate at the lower plenum is made uniform at the inlet of the rotating rack. 4) Thermal-hydraulic behavior in the rotating rack is almost one-dimensional. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-10-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2008-02-15

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

  9. AP1000{sup R} severe accident features and post-Fukushima considerations

    Energy Technology Data Exchange (ETDEWEB)

    Scobel, J. H.; Schulz, T. L.; Williams, M. G. [Westinghouse Electric Company, LLC, 1000 Westinghouse Dr., Cranberry Township, PA 16066 (United States)

    2012-07-01

    The AP1000{sup R} passive nuclear power plant is uniquely equipped to withstand an extended station blackout scenario such as the events following the earthquake and tsunami at Fukushima without compromising core and containment integrity. The AP1000 plant shuts down the reactor, cools the core, containment and spent fuel pool for more than 3 days using passive systems that do not require AC or DC power or operator actions. Following this passive coping period, minimal operator actions are needed to extend the operation of the passive features to 7 days using installed equipment. To provide defense-in-depth for design extension conditions, the AP1000 plant has engineered features that mitigate the effects of core damage. Engineered features retain damaged core debris within the reactor vessel as a key feature. Other aspects of the design protect containment integrity during severe accidents, including unique features of the AP1000 design relative to passive containment cooling with water and air, and hydrogen management. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-03-01

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

  11. Thermochemistry of Ruthenium Oxyhydroxide Species and Their Impact on Volatile Speciations in Severe Nuclear Accident Conditions.

    Science.gov (United States)

    Miradji, Faoulat; Virot, François; Souvi, Sidi; Cantrel, Laurent; Louis, Florent; Vallet, Valérie

    2016-02-01

    Literature thermodynamic data of ruthenium oxyhydroxides reveal large uncertainties in some of the standard enthalpies of formation, motivating the use of high-level relativistic correlated quantum chemical methods to reduce the level of discrepancies. Reaction energies leading to the formation of all possible oxyhydroxide species RuOx(OH)y(H2O)z have been calculated for a series of reactions combining DFT (TPSSh-5%HF) geometries and partition functions, CCSD(T) energies extrapolated to the complete basis set limits. The highly accurate ab initio thermodynamic data were used as input data of thermodynamic equilibrium computations to derive the speciation of gaseous ruthenium species in the temperature, pressure and concentration conditions of severe nuclear accidents occurring in pressurized water reactors. At temperatures lower than 1000 K, gaseous ruthenium tetraoxide is the dominating species, between 1000 and 2000 K ruthenium trioxide becomes preponderant, whereas at higher temperatures gaseous ruthenium oxide, dioxide and even Ru in gaseous phase are formed. Although earlier studies predicted the formation of oxyhydroxides in significant quantities, the use of highly accurate ab initio thermodynamic data for ruthenium gaseous species leads to a more reliable inventory of gaseous ruthenium species in which gaseous oxyhydroxide ruthenium molecules are formed only in negligible amounts. PMID:26789932

  12. Feasibility studies on severe accident mitigation measures against containment over-pressurization

    International Nuclear Information System (INIS)

    The integrity of the containment as the last barrier against radioactive material releases to the environment can be threatened during severe accident. Containment over-pressurization occurs when decay heat and chemical reaction from the degraded core cannot be removed effectively. Then, steam and non-condensable gases are produced by boiling and molten core concrete interaction. If operator actions are not implemented properly, it leads to a containment failure. The applicable means to depressurize the containment includes external cooling of the containment, internal decay heat removal, and containment venting. So, the feasibility study on containment backup spray system and containment filtered venting system are examined on PWR with large dry containment. Thermo-hydraulic analysis results show that both options can effectively de-pressurize the containment against representative containment over-pressurization scenarios. Containment backup spray system can prevent radiological release and maintain containment integrity as an active way. But optimal operating strategy is needed to prevent flooding inside the containment and hydrogen flammability. Filtered venting can mitigate radiological release below 10-3 of initial inventory as a passive way. But the molten core concrete interaction cannot be mitigated. Overall, each option has its own pros. and cons. So, the final design option should be chosen considering the overall safety effect based on thermo-hydraulic analysis, risk reduction factor, incorporation into design concept (diversity .vs redundancy), reliability of each component (passive .vs active), operability, etc. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

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

    International Nuclear Information System (INIS)

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

  15. A study on the pressurized water reactor (PWR) containment response analysis methodologies for postulated severe accident

    International Nuclear Information System (INIS)

    The present study contains two major parts: one is the treatment of uncertainties involved in the current APET and the other is the importance analysis of the APET uncertainty inputs. A clear disadvantage of the expert opinion polling process approach for uncertainty analysis of the current probabilistic risk assessment (PRA) is that the sufficient robustness in the final results may not be attained against the ambiguity of the information upon which the experts base their judgement or the judgmental uncertainty arising under various imprecise and incomplete information. For the treatment of such type of uncertainty, a new approach based on fuzzy set theory is proposed. Then its potential use to the uncertainty analysis of the current PRA is proved through an analysis of accident progression event tree (APET). As a product, a formal procedure with computational algorithms suitable for application of the fuzzy set theory to the APET analysis is provided. Comparing with the uncertainty analysis results obtained by the statistical approach currently used in PRA, the present approach has several major advantages: Firstly, it greatly enhances the robustness in the final results of APET uncertainty analysis by modeling the judgmental uncertainty that arises in the probabilistic quantification of APET top events. Secondly, the modeling of APET uncertainty analysis is far more convenient because of the nonprobabilistic features of fuzzy probabilities used for uncertainty quantification of the APET top events. Thirdly, the APET model can easily be operated by means of a well defined formal propagation logic of fuzzy set theory without going through a tedious sampling procedure. Finally, the fuzzy outcomes provide at least as much information as the existing methods based on the statistical approach. Thus, the present approach can be used as a valuable alternative approach to uncertainty analysis used in the current PRA. Two importance measures for the importance analysis of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  17. ALIBABA: a French Expert System for PWR Containment Analysis in case of Severe Accidents

    International Nuclear Information System (INIS)

    In the event of an accident occurring in a French pressurized water reactor (PWR), the authorities should be in position to implement the measures required to protect the surrounding population and the environment from radiological consequences of potential releases. The Institute for Nuclear Safety and Protection is part of the national emergency organization established for this purpose. It provides technical support to the French nuclear safety authority. As a technical adviser, IPSN has defined a methodology intended to help assess the plant status and monitor its development as soon as the accident is detected. On the basis of this assessment, the method forecasts the potential behavior of the installation and estimates the related consequences. The state of the installation is evaluated throughout the accident with special reference to the three barriers stretched out between the radioisotopes and the environment (fuel cladding, reactor coolant system and the containment building). It considers successively their physical state, the state of the safety functions guaranteeing their integrity and finally the state of the systems available to monitor these functions. In order to properly diagnose and predict the state of the barriers, evaluations are necessary to quantify parameters such as the break size on the reactor coolant system, the time to core uncovering and the core degradation. As a result, fission products behavior inside the installation and releases outside the plant are assessed. Several flexible, rapid and user-friendly software tools, which are part of the French SESAME system, have been developed to help the experts with their assessment. products cannot be realistically quantified without a complete knowledge of the state of the containment barrier. The expert system ALIBABA is separate from these tools. It provides complementary qualitative information about the third barrier. Indeed, the ongoing or potential releases of fission products

  18. LWR severe accident simulation: Iodine behaviour in FPT2 experiment and advances on containment iodine chemistry

    International Nuclear Information System (INIS)

    Highlights: ► Short term gaseous iodine fraction can be produced either in primary circuit or on containment condensing surfaces. ► Gaseous radiolytic reactions convert volatile iodine into non-volatile iodine oxide particulates. ► Alkaline and evaporating sump decrease the iodine volatility in containment. ► Release of volatile iodine from containment surfaces explained the long term stationary residual gaseous iodine concentration. - Abstract: The Phebus Fission Product (FP) Program studies key phenomena of severe accidents in water-cooled nuclear reactors. In the framework of the Phebus program, five in-pile experiments have been performed that cover fuel rod degradation and behaviour of fission products released via the coolant circuit into the containment vessel. The focus of this paper is on iodine behaviour during the Phebus FPT2 test. FPT2 used a 33 GWd/t uranium dioxide fuel enriched to 4.5%, re-irradiated in situ for 7 days to a burn-up of 130 MWd/t. This test was performed to study the impact of steam-poor conditions and boric acid on the fission product chemistry. For the containment vessel, more specifically, the objective was to study iodine chemistry in an alkaline sump under evaporating conditions. The iodine results of the Phebus FPT2 test confirmed many of the essential features of iodine behaviour in the containment vessel provided by the first two Phebus tests, FPT0 and FPT1. These are the existence of an early gaseous iodine fraction, the persistence of low gaseous iodine concentrations and the importance of the sump in suppressing the iodine partitioning from sump to atmosphere. The main new insights provided by the Phebus FPT2 test were the iodine desorption from stainless steel walls deposits and the role of the evaporating sump in further iodine depletion in the containment atmosphere. The current paper presents an interpretation of the iodine behaviour in the FPT2 containment vessel based on dedicated small-scale analytical

  19. Analysis of severe accident on OPR1000 PWR plant at low power and shutdown states with MAAP5 code

    International Nuclear Information System (INIS)

    The objective of this paper is to provide a brief description of severe accident analysis using computer codes in Korean OPR1000 Plant at low power and shutdown states. The results of the analysis are utilized in preparing the shutdown severe accident management guidelines (LPSD SAMG). As part of the efforts to prepare LPSD SAMG, analysis of severe accident is performed at low power and shutdown states with MAAP5 code. The Korean OPR1000 plant, a PWR plant with 2 hot legs and 4 cold legs is considered as a reference plant in the analysis. In this study, the scenarios are selected based on the plant operational states (POS) and dominant initiating events (IE) which cause the core damages. Typical scenarios are the loss of shutdown cooling (LSCS) at various primary coolant levels and stuck-opening of valves which prevent the low temperature over pressurization (LTOP) of primary system. As the analysis results, the core uncovery is expected in 2∼6 hours. The maximum temperature of core exit exceeds 649degC (SAMG entry temperature) in 3∼7 hours. The molten corium starts to relocate into lower head in 5∼13 hours and reactor vessel failure is occurred in 11∼14 hours. The above mentioned timings are utilized to choose the possible actions and the timing to implement those actions LPSD SAMG. Also based on the results, the environmental conditions that instruments may encounter in a severe accident are determined. (author)

  20. Summary of severe accident assessment for Atucha 2 Nuclear Power Plant using RELAP5/SCDAPSIM Mod3.6

    International Nuclear Information System (INIS)

    A severe accident assessment was performed for the Atucha 2 Nuclear Power Plant in Argentina. Atucha 2 is a PHWR, cooled and moderated by heavy water, presently in commissioning process. Its 451 fuel assemblies are 6.03m high and each composed of 37 Zircaloy clad fuel rods. Each assembly is placed inside an individual Zircaloy coolant channel. Heavy water coolant flows inside the channels which are all immersed inside the moderator tank. The RPV lower plenum is occupied by a massive steel structure called 'filling body' that was designed to minimize heavy water inventory. Due to some unique design characteristics, severe accident progression in Atucha 2 is expected to be somewhat different from that predicted for regular PWRs. Therefore, a very detailed assessment was performed, focused on the different accident stages and expected phenomena by the use of different input models and nodalizations. When possible, linking to available experimental data was performed. RELAP/SCDAPSIM Mod 3.6 was the computer code selected to perform this task. The modeling of Atucha 2's unique characteristics required several extensions to the code. For the severe accident assessment of Atucha 2, three different input models were developed that were key instruments for the debugging and evaluation process. A Single Channel Model was used to evaluate the first stages of core heatup (including the boiloff of the channels and moderator tank), an RPV standalone model was used to assess the interaction between components in the complete core and for the evaluation of late in-core melting and relocation. Then, a Lower Plenum standalone model was developed to assess the behavior of the melted and slumped core material on top of the filling body and to analyze ex-vessel cooling as a possible severe accident management action. For each of the cases, highlights of key results are shown and general conclusions are drawn. In the case of a severe accident with significant meltdown of

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-07-01

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

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

    Science.gov (United States)

    Petrov, A.; Petrova, D.

    2016-08-01

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

  3. Proceedings of the European Review Meeting on Severe Accident Research - ERMSAR 2005

    International Nuclear Information System (INIS)

    The SARNET network has been set up under the aegis of the Framework Programmes (FP) of the European Commission on research. Two projects have been defined, both coordinated by IRSN (France), in the FP6 (2004-08) and FP7 (2009-13), with the following key objectives: Improving knowledge on severe accidents (SA) in order to reduce the uncertainties on the pending issues, thereby enhancing the plant safety, Coordinating research resources and expertise available in Europe, Preserving the research data and disseminating knowledge. The network members commit to contribute to a Joint Programme of Activities that can be broken into several elements: - Implementing an advanced communication tool for fostering exchange of information; - Harmonizing and re-orienting the research programmes, and defining commonly new ones; - Analysing commonly the experimental results provided by research programmes in order to elaborate a common understanding of concerned phenomena; - Developing ASTEC, which capitalizes in terms of models the knowledge produced within SARNET; - Developing Scientific Databases, in which all the results of research programmes are stored; - Developing a common methodology for Probabilistic Safety Assessment (PSA) of NNPs; - Developing educational courses and text (source) books; - Promoting personnel mobility between the various European organisations. SARNET provides an appropriate frame for achieving within a couple of years a sustainable integration of the European research capacities on SA. By capitalizing the acquired knowledge in ASTEC and in Scientific Databases, SARNET produces necessary conditions for preserving the knowledge produced by thousands of men-years and diffusing it to a large number of end-users. By fostering collaborative work on developing and validating ASTEC, SARNET makes this code as the European reference for any kind of water-cooled NPP existing in Europe. By fostering collaborative work in the domain of code development and PSA

  4. Mitigative techniques and analysis of generic site conditions for ground-water contamination associated with severe accidents

    International Nuclear Information System (INIS)

    The purpose of this study is to evaluate the feasibility of using ground-water contaminant mitigation techniques to control radionuclide migration following a severe commercial nuclear power reactor accident. The two types of severe commercial reactor accidents investigated are: (1) containment basemat penetration of core melt debris which slowly cools and leaches radionuclides to the subsurface environment, and (2) containment basemat penetration of sump water without full penetration of the core mass. Six generic hydrogeologic site classifications are developed from an evaluation of reported data pertaining to the hydrogeologic properties of all existing and proposed commercial reactor sites. One-dimensional radionuclide transport analyses are conducted on each of the individual reactor sites to determine the generic characteristics of a radionuclide discharge to an accessible environment. Ground-water contaminant mitigation techniques that may be suitable, depending on specific site and accident conditions, for severe power plant accidents are identified and evaluated. Feasible mitigative techniques and associated constraints on feasibility are determined for each of the six hydrogeologic site classifications. The first of three case studies is conducted on a site located on the Texas Gulf Coastal Plain. Mitigative strategies are evaluated for their impact on contaminant transport and results show that the techniques evaluated significantly increased ground-water travel times. 31 references, 118 figures, 62 tables

  5. Assessment of in-vessel corium retention by external reactor vessel cooling during severe accidents in CPR1000

    International Nuclear Information System (INIS)

    The In-Vessel corium Retention (IVR) through External Reactor Vessel Cooling (ERVC) during Severe Accidents in a Chinese 1000 MW class commercial pressurized water reactor (CPR1000) is assessed by Risk-Oriented Accident Analysis Methodology (ROAAM), which was developed specifically for the purpose of addressing rare and high consequence hazard issues and had been applied to the IVR studies in various Nuclear Power Plants (NPPs) worldwide. IVR is considered as one of promising strategies for severe accidents mitigation, and CPR1000 will be a major option for the construction of NPPs at the next decades in China. Hence, applicability study of IVR in CPR1000 is drawing great attention and interest. This is almost the first IVR study in Chinese NPPs. Four representative severe accident scenarios are selected for the IVR assessment in the paper. The core-melting and relocation progress conditions of these scenarios are calculated through MAAP code. According to the MAAP calculation combined with engineering judgment, the input probability distribution of the assessment is determined. Success probability of IVR-ERVC from the viewpoint of thermal failure is then predicted using ROAAM. MOPOL, a code for the calculation of the heat flux distributions of the reactor vessel outer wall in severe accidents condition, is developed for the performance of this IVR assessment. It is demonstrated that IVR-ERVC itself is fairly effective for reactor vessel integrity. Application of IVR-ERVC strategy in CPR1000 is envisioned probable if a further more comprehensive risk-benefit evaluation conclusion is positive. (author)

  6. An experimental study on layer inversion in the corium pool during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Kyoung-Ho, E-mail: khkang@kaeri.re.kr; Park, Rae-Joon; Hong, Seong-Ho; Hong, Seong-Wan; Ha, Kwang Soon

    2014-10-15

    Highlight: • COSMOS tests were performed to investigate the layer inversion of the corium pool using prototypic materials. • An induction heating method was implemented for melting of the UO{sub 2}–ZrO{sub 2}–Zr–Fe mixture. • The simulated sequence was the TLFW in APR1400. • A metallurgical inspection was performed to investigate the layer inversion. • COSMOS test results show the possibility of layer inversion of the heavy metallic material. - Abstract: COSMOS (COrium configuration of the molten State in the MOst Severe accidents) tests using prototypic materials have been performed to investigate the layer inversion of the heavy metallic material in the corium pool. An induction heating method using a cold crucible was implemented for melting of the UO{sub 2}–ZrO{sub 2}–Zr–Fe mixture. Before the main test, three preliminary tests have been performed to enhance the experimental techniques using the real core material. One main test of the COSMOS has been performed to evaluate the corium pool configuration in the lower plenum of the reactor vessel for the TLFW (Total Loss of Feed Water) sequence of the APR (Advanced Power Reactor) 1400 under the IVR-ERVC (In-Vessel corium Retention through External Reactor Vessel Cooling). A post-test examination of cutting, EPMA, and XRD for the solidified corium pool ingot in the main test has been performed to investigate the layer inversion. From the three preliminary tests, melting techniques of the real core material using a cold crucible were successfully developed. The metallurgical inspection results on the chemical information coincide with the visual observation on the centerline cut ingot in that the upper part is metal and the lower lump is an oxidic mixture with some metal clods in the main test. This means the possibility of layer inversion of the heavy metallic material in the corium pool.

  7. Radionuclide releases from UO2 and MOX fuel under severe accident conditions

    International Nuclear Information System (INIS)

    Radionuclide release from fuel under severe accident conditions was investigated in VEGA (Verification Experiments of radionuclides Gas/Aerosol release) program at Japan Atomic Energy Agency (JAEA). This study compares the results of tests on PWR-UO2 fuel, BWR-UO2 fuel and ATR (Advanced Thermal Reactor)-MOX(mixed oxide) fuel. The three types of fuels have burnup of 47, 56 and 43 GWd/t, respectively. Each fuel without cladding was set in a tungsten crucible and heated up to about 3130 K in helium atmosphere at 0.1 MPa. The fuel temperature was kept constant for 10 to 20 minuets at four plateaus during the heat up. The total fractional releases of high volatile Cs were 100% for the PWR-UO2 fuel, 97% for the BWR-UO2 fuel and 97% for the ART-MOX fuel. The Cs release with the heatup was different among three fuels for the temperature range below 2310 K, while the difference became small for the higher temperature range. The difference for the lower temperature range is considered to be caused by difference of irradiation histories, which varies migration states of the high volatile element. The total fractional releases of Mo and U were in the order of 0.1% and those of Sr and Pu were in the order of 1% both the tests with the BWR-UO2 and the ATR-MOX fuels. Release of low volatiles, U, Pu, Sr and Mo were dependent strongly on their chemical states, suggesting that vaporization was the controlling process. Namely, release of Pu and Sr was enhanced by the reduction of oxide, while it was largely decreased for Mo even at higher temperatures in the same atmosphere. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  9. Critical heat flux for APR1400 lower head vessel during a severe accident

    International Nuclear Information System (INIS)

    Highlights: ► Studied boiling on downward-facing hemispherical vessel with asymmetric thermal insulator. ► Scaled the APR1400 lower head linearly down by 1/10 including ICI tubes and shear keys. ► Performed thermal analysis using ANSYS V11.0 to determine the internal temperature and heat flux. ► Performed tests to obtain the CHF with saturated demineralized water at atmospheric pressure. ► Measured CHF accounting for 3D random flow effect expected in the APR1400 application. -- Abstract: Corium Ablation Stopper Apparatus (CASA) has a downward-facing hemispherical vessel and geometrically asymmetric thermal insulator of the Advanced Power Reactor 1400 MWe (APR1400) scaled linearly down by 1/10, as well as sixty-one in-core instrumentation (ICI) tubes and four shear keys. The heated vessel plays a pivotal role in CASA depending on the configuration of the oxide pool and metal layer to bring about the focusing effect expected of a molten pool in the lower head during a severe accident. The heated vessel was designed through a trial-and-error method and thermal analysis. Thermal analysis was performed using ANSYS V11.0 to investigate the effect of the internal temperature and heat flux on the integral hemispherical copper vessel. The CASA tests were carried out to obtain the critical heat flux (CHF) with saturated and demineralized water at the atmospheric pressure (0.1 MPa). The CHF in the metal layer through the hemispherical channel was found to be lower than that in the ULPU-2400 configuration V data through the streamlined thermal insulator. The experimental CHF was measured and obtained through the CASA hemispherical heated surface accounting for the three-dimensional random flow effect expected in the APR1400 application

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

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

    International Nuclear Information System (INIS)

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

  12. Experimental study on in-vessel debris coolability during severe accident

    International Nuclear Information System (INIS)

    A research program, called SONATA-IV(Simulation of Naturally Arrested Thermal Attack In-Vessel), has been performed to verify the gap cooling mechanism of corium in the lower plenum, and to develop management and mitigation strategies under severe accident conditions. For the proof-of-principles experiment, the LAVA(Lower-plenum Arrested Vessel Attack) experiments have been performed to gather proof of gap formation and to evaluate the gap effect on in-vessel cooling, using Al2O3/Fe (or Al2O3 only) thermite melt as corium simulant. And also the CHFG(Critical Heat Flux in Gap) experiments have been performed to measure the critical power and to investigate the inherent cooling mechanism in the hemispherical narrow gap. In addition to the experiments, LILAC code was developed to analyze and predict the thermo-hydraulic phenomena of the corium relocated in the reactor lower plenum. It could be found from the LAVA and CHFG experimental results that continuous gap ranged from 1 to 5 mm was formed and that maximum heat removal capacity through a gap is a key factor in determining the potentials of the integrity of the vessel. After all the possibility of IVR(In-Vessel corium Retention) through gap cooling highly depends on the melt relocated into the lower plenum and the gap size. So, feasibility experiments have been performed for the assessment of improved IVR concepts using an internal engineered gap device and a dual strategy of In/Ex-vessel cooling using the LAVA facility. It is preliminarily concluded that these cooling measures lead to an enhanced cooling of the corium in the lower plenum of the reactor vessel. The additional studies will be performed to verify the quantitative heat removal capacity for these cooling measures in the 2nd phase of mid- and long term project period

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-08-01

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

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

    International Nuclear Information System (INIS)

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

  15. Modelling of cladding oxidation by air under severe accident conditions with the MAAP 4 code

    International Nuclear Information System (INIS)

    In a nuclear power plant, air ingress into the vessel is a potential risk in some low probable situations of severe accidents. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of FP. This is particularly true speaking about ruthenium release, which can be significantly increased in the presence of air. This is a key issue due to the high radio-toxicity of ruthenium and its ability to form highly volatile oxides. The oxygen affinity is decreasing in priority from the Zircaloy cladding, to fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. As a first step, a phenomenological study has been carried out to characterize nitriding of the Zircaloy claddings. In summary, nitriding occurs preferentially when the oxygen has been consumed locally or in case of total oxygen starvation and when the cladding was slightly pre-oxidized. Just like oxidation, nitriding can be modeled in a simplified form as a cladding weight gain in terms of thickness. The model implemented in MAAP takes this into account as well as re-oxidation of the nitrides, in the case where oxygen is available again (especially during a reflood). Several correlations were thus integrated and a new one, called “KIT-EDF”, was developed, based on KIT separate-effect tests. The model has been implemented and validated against QUENCH-16 and QUENCH-10 experiments, studying the oxidation in air atmosphere of an assembly pre-oxidized in steam and finally quenched with water. The simulations give encouraging results since the modeling of nitriding effects has increased hydrogen production during reflood, as experimentally observed. The results of this study lead us to identify a number of perspectives for the future, namely taking into account the changes in the structure of the oxide layer during a

  16. Studies on reactor piping integrity during severe accident in WIND project

    International Nuclear Information System (INIS)

    In a severe accident of a light water reactor, the reactor coolant piping might be subjected to thermal loads resulted from decay heat release of the deposited fission products, and heat transfer from high temperature gases generated in the reactor core region in addition to an internal pressure load. Thermal and structural responses of the reactor coolant piping under elevated temperature and pressure conditions are being investigated in piping integrity tests in WIND (Wide Range Piping Integrity Demonstration) project at Japan Atomic Energy Research Institute. Four piping failure tests have been conducted using straight stainless steel pipes with a diameter of 114.3mm or 355.6mm which simulate a part of the reactor coolant piping. In a recent failure test in which a pipe with an outer diameter of 355.6mm and a wall thickness of 35.7mm was used and an internal pressure was kept at 10MPa, a failure of the pipe was observed when the maximum temperature of the pipe reached 1040degC after the temperature was sustained at 1000degC for about one hour. Several small openings were formed at the pipe failure in an area showing the maximum temperature. A history of the change in a vertical diameter was obtained from the measurement of displacement at the top and the bottom of the pipe at the center of heated area. An acceleration of piping ballooning was observed when the maximum temperature of the pipe became approximately 1000degC. It was confirmed in post-test observation that the pipe diameter in vertical and horizontal direction was remarkably enlarged and the pipe wall thickness was largely reduced. In parallel with the tests, a post-test analysis was performed using ABAQUS code. The results from 2D elasto-plastic creep analysis for a piping failure test with an outer diameter of 114.3mm and a wall thickness of 13.5mm qualitatively reproduced the final deformation including the diameter enlargement and the wall thinning. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-05-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Cave, L.; Min, T.K.

    1980-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-05-01

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

  3. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Volume 6, Part 2: Appendices

    International Nuclear Information System (INIS)

    The objectives are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed. Volume 1 summarizes the results of the study. The scope of the level-1 study includes plant damage state analyses, and uncertainty analysis. The internal event analysis is documented in Volume 2. The internal fire and internal flood analysis are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc. A phased approach was used in the level 2/3 PRA program, however both phases addressed the risk from only mid-loop operation. The first phase of the level 2/3 PRA was initiated in late 1991 and consisted of an Abridged Risk Study. This study was completed in May 1992 and was focused on accident progression and consequences, conditional on core damage. Phase 2 is a more detailed study in which an evaluation of risk during mid-loop operation was performed. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6. This report, Volume 6, Part 2, consists of five appendices containing supporting information for: the PDS (plant damage state) analysis; the accident progression analysis; the source term analysis; the consequence analysis; and the Melcor analysis. 73 figs., 21 tabs

  4. PARIS project: Radiolytic oxidation of molecular iodine in containment during a nuclear reactor severe accident

    International Nuclear Information System (INIS)

    In case of a hypothetical severe accident in a nuclear LWR (light water reactor), the high radiation fields reached in the reactor containment building due to the release of fission products from the reactor core could induce air radiolysis. The air radiolysis products could, in turn, oxidise gaseous molecular iodine into aerosol-borne iodine-oxygen-nitrogen compounds. Thereby, this reaction involves a change of iodine speciation and a decrease of iodine volatility in the reactor containment atmosphere. Kinetic data were produced within the PARIS project on the air radiolysis products formation and destruction, and on their reaction with molecular iodine, with the objective of developing and validating existing kinetic models. The current paper includes the non-iodine tests of the PARIS project whose objective was to determine the rates of formation and destruction of air radiolysis products in the presence of both structural containment surfaces (decontamination coating ('paint') and stainless steel), aerosol particles such as silver rich particles (issued from the control rods) in boundary conditions representative for LWR or PHEBUS facility containments. It is found that the air radiolysis products concentration increases with dose and tend to approach saturation levels at doses higher than about 1 kGy. This behaviour is more evident in oxygen/steam atmospheres, producing ozone, than in air/30% (v/v) steam atmospheres, the latter favouring the model-predicted on-going production of nitrogen dioxide even at very high doses. No significant effect of temperature, dose rate and hydrogen addition (4%, v/v) was observed. Furthermore, the inserted surfaces do not exhibit significant effects on the air radiolysis concentrations. However, these 'non-noticeable influence' could be due to a masking of small effects by the appreciable scattering of the experimental air radiolysis product concentrations. The PARIS results are then analysed using two different kinetic models

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-02-01

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

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

    International Nuclear Information System (INIS)

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

  9. Fuel Behaviour and Modelling under Severe Transient and Loss of Coolant Accident (LOCA) Conditions. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    In recent years the demands on 'fuel duties' have increased, including transient regimes, higher burnups and longer fuel cycles. To satisfy these demands, fuel vendors have developed and introduced new cladding and fuel material designs to provide sufficient margins for safe operation of the fuel components. National and international experimental programmes have been launched, and models have been developed or adapted to take into account the changed conditions. These developments enable water cooled reactors, which contribute about 95% of the nuclear power in the world today, to operate safely under all operating conditions; moreover, even under severe transient or accident conditions, such as reactivity initiated accidents (RIAs) or loss of coolant accidents (LOCAs), the behaviour of the fuel can be adequately predicted and the consequences of such events can be safely contained. In 2010 the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) recommended that a technical meeting on ''Fuel Behaviour and Modelling under Severe Transient and LOCA Conditions'' be held in Japan. The accident at the Fukushima Daiichi nuclear power plant in March 2011 highlighted the need to address this subject, and despite the difficult situation in Japan at the time, the recommended plan was confirmed, and the Japan Atomic Energy Agency (JAEA) hosted the technical meeting in Mito, Ibaraki Prefecture, Japan, from 18 to 21 October 2011. This meeting was the eighth in a series of IAEA meetings, which reflects Member States' continuing interest in the above issues. The previous meetings were held in 1980 (jointly with OECD Nuclear Energy Agency, Helsinki, Finland), 1983 (Riso, Denmark), 1986 (Vienna, Austria), 1988 (Preston, United Kingdom), 1992 (Pembroke, Canada), 1995 (Dimitrovgrad, Russian Federation) and 2001 (Halden, Norway). The purpose of the technical meeting was to provide a forum for international experts to review the current situation and the state of

  10. Evaluation of thermal behaviour of zirconia ceramics plane core catcher at severe accident of the nuclear reactor

    International Nuclear Information System (INIS)

    Physical and mathematical models were constructed and prompt average prediction of the thermal state of the zirconium dioxide ceramics plane core catcher under severe accident of nuclear reactor was assumed. Essential effect of radiation transfer on the dynamics of thermal processes in the catcher was shown in the concept of approximation of radiation thermal conduction. Effect of the partial impregnation of the catcher bottom by molten iron oxides on quasistationary state was investigated

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-08-01

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

  13. The mechanical behaviour of pressurized water reactor containments during severe accidents

    International Nuclear Information System (INIS)

    The programme of investigations of the behaviour of the safety containment during the build-up of pressure during a postulated core melt-out accident and the behaviour of the safety containment if there is vibration caused by a serious earthquake is introduced. The first results of theoretical and experimental investigations of pressure to cause failure and the type of failure of the safety containment in a core melt-out accident are discussed. Calculations showed that the safety containment failed at an inside excess pressure of about 14 bar in the transition zone to the thicker rings due to plastic instability. Further investigation will have to make clear whether an earlier failure at another weak place will prevent such a great rise of pressure. (orig.)

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

    OpenAIRE

    Arndt, J; Grebner, H.; J. Sievers

    2012-01-01

    During postulated high-pressure core melt accident scenarios, temperature values of more than 800°C can be reached in the reactor coolant line and the surge line of a pressurised water reactor (PWR), before the bottom of the reactor pressure vessel experiences a significant temperature increase due to core melting. For the assessment of components of the primary cooling circuit, two methods are used by GRS. One is the simplified method ASTOR (approximated structural time of rupture). This met...

  15. Resuspension of fission products during severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    This report investigates the influence of resuspension phenomena on the overall radiological source term of core melt accidents in a pressurized water reactor. A review of the existing literature is given and the literature data are applied to calculations of the source term. A large scatter in the existing data was found. Depending on the scenario and on the data set chosen for the calculations the relative influence of resuspended fission products on the source term ranges from dominant to negligible. (orig.)

  16. Precursors to potential severe core damage accidents: 1992, a status report

    International Nuclear Information System (INIS)

    This document is part of a report which documents 1992 operational events selected as accident sequence precursors. This report describes the 27 precursors identified from the 1992 licensee event reports. It also describe containment-related events; open-quote interesting close-quote events; potentially significant events that were considered impractical to analyze; copies of the licensee event reports which were cited in the cases above; and comments from the licensee and NRC in response to the preliminary reports

  17. Phenomenological Studies on Melt-Structure-Water Interactions (MSWI) during Postulated Severe Accidents

    International Nuclear Information System (INIS)

    This is the annual report for the work performed in year 2003 in the research project 'Melt-Structure-Water Interactions (MSWI) During Severe Accidents in LWRs', under the auspices of the APRI Project, jointly funded by SKI, HSK, and the Swedish and Finnish power companies. The emphasis of the work was placed on phenomena and parameters, which govern the droplet fragmentation in steam explosions, in-vessel and ex-vessel melt/debris coolability, melt pool convection, and the thermal and mechanical loadings of a pressure vessel during melt-vessel interaction. Most research projects in 2002, such as the COMECO, POMECO and MISTEE programs, were continued. An analysis of the FOREVER experiments using the RELAP code to investigate the melt coolability, bubble dynamics and bubble stability to investigate the dynamic behavior of vapor bubble during steam explosions and associated melt fragmentation, quenching boiling experiment to investigate the thermal behavior of single melt droplet were newly initiated. The SIMECO experiment to investigate the three-layer melt pool convection was restarted. The experimental facilities for these projects were fully functional during year 2003. Many of the investigations performed during the course of the MSWI project have produced papers, which have been published in the proceedings of technical meetings and Journals. Significant technical advances were achieved during the course of these studies. These were: A series of experiments on single drop steam explosions was performed to investigate the fine fragmentation process of a metallic melt drop in various thermal conditions. For the first time, transient fine fragmentation process of a melt drop during explosion phase of a steam explosion was visualized continuously and quantified. Different triggering behavior with respect to the coolant subcooling was observed. The analyses on bubble dynamics during a single drop steam explosion and vapor bubble stability estimated the dynamic

  18. Pilot program: NRC severe reactor accident incident response training manual: Public protective actions: Predetermined criteria and initial actions

    International Nuclear Information System (INIS)

    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. Public Protective Actions - Predetermined Criteria and Initial Actions is the fourth 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 reviews public protective action criteria and objectives, their bases and implementation, and the expected public response. Each volume serves, respectively, as the text for a course of instruction in a series of courses for NRC response personnel. These materials do not provide guidance or license requirements for NRC licensees. 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. Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

    1997-07-01

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

  20. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

  1. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    International Nuclear Information System (INIS)

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

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

    Directory of Open Access Journals (Sweden)

    Kil-Mo Koo

    2012-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    BLOWER,DANIEL F.; CLAUSS,DAVID B.

    1999-10-01

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

  5. Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors

    Energy Technology Data Exchange (ETDEWEB)

    Spencer, B.W.; Marchaterre, J.F.; Anderson, R.P.; Armstrong, D.R.; Baker, L.; Cho, D.H.; Gabor, J.D.; Pedersen, D.R.; Sienicki, J.J.; Stein, R.P.

    1986-01-01

    Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO/sub 2//sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included.

  6. Status of ANL out-of-pile investigations of severe accident phenomena for liquid metal reactors

    International Nuclear Information System (INIS)

    Research addressing LMFBR whole core accidents has been terminated, and there is now emphasis on quantifying reactivity feedbacks, and in particular enhancing negative feedback, so that advanced LMR designs will provide inherently safe operation. The status of recent HCDA-related laboratory research performed at ANL, up to the time that such activities were no longer needed to support CRBR licensing, is described. Included are descriptions of programs addressing sodium channel voiding, fuel sweepout, fuel dispersal and plugging, boiled-up pool, UO2/sodium FCI, and debris coolability. Descriptions of recent investigations involving the metal fuel/sodium system are also included

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-03-15

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

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

    Science.gov (United States)

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

    2016-09-01

    Purpose Case management is widely accepted as an effective method to support medical rehabilitation and vocational reintegration of accident victims with musculoskeletal injuries. This study investigates whether more intensive case management improves outcomes such as work incapacity and treatment costs for severely injured patients. Methods 8,050 patients were randomly allocated either to standard case management (SCM, administered by claims specialists) or intensive case management (ICM, administered by case managers). These study groups differ mainly by caseload, which was approximately 100 cases in SCM and 35 in ICM. The setting is equivalent to a prospective randomized controlled trial. A 6-year follow-up period was chosen in order to encompass both short-term insurance benefits and permanent disability costs. All data were extracted from administrative insurance databases. Results Average work incapacity over the 6-year follow-up, including contributions from daily allowances and permanent losses from disability, was slightly but insignificantly higher under ICM than under SCM (21.6 vs. 21.3 % of pre-accident work capacity). Remaining work incapacity after 6 years of follow-up showed no difference between ICM and SCM (8.9 vs. 8.8 % of pre-accident work incapacity). Treatment costs were 43,500 Swiss Francs (CHF) in ICM compared to 39,800 in SCM (+9.4 %, p = 0.01). The number of care providers involved in ICM was 10.5 compared to 10.0 in ICM (+5.0 %, p reintegration of accident victims. PMID:26687330

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

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

  13. Use of open source software in estimating the effects of a severe accident on the Mark II containment

    International Nuclear Information System (INIS)

    Because the spectrum of scenarios of severe accident before which must verify the integrity of the containment can be very broad, it arises here a calculation methodology to estimate the structural response of the containment without incurring in high costs for commercial software licenses, or in times and calculation excessive requirements. The capabilities of computer programs with license of open source, OpenFOAM for CFD calculations and Salome-Meca for thermal and mechanical calculations were tested. The methodology begins of the venting of mass and energy that are postulated inside the container and the values of the thermal and mechanical fields are obtained through the walls. (Author)

  14. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  15. Fission product chemistry in severe nuclear reactor accidents, specialists' meeting at JRC-Ispra, 15-17 January 1990

    International