WorldWideScience

Sample records for blackout severe accident

  1. Management of severe accidents in Indian PHWRs with special reference to station blackout

    International Nuclear Information System (INIS)

    Subramani, V.A.; Rajasabai, N.; Murthy, K.S.N.; Singh, S.P.

    1988-01-01

    The paper describes the approach adopted in India to severe accidents with special reference to failure of on-site and off-site power to PHWR plants. The basic design philosophy followed in control and protection systems is explained with reference to core heat removal. An introduction is given to a station electrical service system which is divided into four categories depending on the reliability called for by the equipment being serviced. Failures leading to a blackout condition and effects on the reactor (calandria) components and end shields are discussed. Operational procedures to reduce the consequences of a blackout on reactor components are enumerated. The various protection systems are considered along with their stand-by systems, especially with respect to core heat removal. For the unlikely event of a station blackout coupled with multiple levels of safety system failures leading to small external release of activity, emergency response plans which include the participation of local, district and state government authorities for shelter and evacuation have been developed. A systematic method of assessing any safety related occurrence is carried out by multi-tiered committees starting from the Unit Safety Committee to the Atomic Energy Regulatory Board. (author)

  2. Core Cooling Assessment of SMART against Severe Station Blackout Accident Scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Won Jae; Kim, Hark Rho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-05-15

    Recent Fukushima disaster was caused by a complete loss of electricity, that is, station blackout followed by unpredicted earthquake and consequent tsunami. This necessitated a re-examination of nuclear plant safety against station blackout accident scenarios. System-integrated Modular Advanced ReacTor (SMART) is an integrated pressurized water reactor developed by KAERI, whose standard design is under regulatory review by KINS. Intrinsic safety of the SMART is featured by: elimination of large pipe breaks, passive residual heat removal, large coolant inventory, low power density, high secondary design pressure and large containment, etc. Unlike Fukushima Mark-I design, SMART passive safety is insured by four-train passive residual heat removal system (PRHRS) that provides natural circulation cooling in the secondary sides of steam generators. In addition, two emergency diesel generators (DG) and an alternative diesel generator insure the AC power supply to active engineered safety features and twelve passive auto-catalytic recombiners in containment prevents potential hydrogen explosion. Thus, it is quite unlikely for SMART to experience Fukushima consequences. Nevertheless, it is worthwhile to assess SMART safety for severe station blackout scenarios in which multiple failures of DGs and PRHRSs are postulated. Thermal hydraulic response of the SMART system is assessed using a best-estimate code, MARS3.1 in order to realistically estimate the time afforded for operator's mitigation actions

  3. Blackouts

    Science.gov (United States)

    Blackouts Before the blackout... - If you have an electric garage door opener, locate the manual release lever and learn how to operate it. - Keep your car’s gas tank at least half full; gas stations ...

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

    OpenAIRE

    Prošek, Andrej; Cizelj, Leon

    2013-01-01

    Stress tests performed in Europe after accident at Fukushima Daiichi also required evaluation of the consequences of loss of safety functions due to station blackout (SBO). Long-term SBO in a pressurized water reactor (PWR) leads to severe accident sequences, assuming that existing plant means (systems, equipment, and procedures) are used for accident mitigation. Therefore the main objective was to study the accident management strategies for SBO scenarios (with different reactor coolant pump...

  5. Accident management for severe accidents

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Zhao, Haihua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Hongbin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zou, Ling [Idaho National Lab. (INL), Idaho Falls, ID (United States); Anders, David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Martineau, Richard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2014-10-01

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

  7. Severe accident phenomena

    International Nuclear Information System (INIS)

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

    1995-02-01

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

  8. Accident Analysis of Chinese CPR1000 in Response to Station Blackout

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Juyoul [FNC Technology Co., Yongin (Korea, Republic of); Cilliers, Anthonie [North-West University, Potchefstroom (South Africa)

    2016-10-15

    Stress tests required evaluation of the consequences of loss of safety functions from any initiating event (e.g., earthquake or flooding) causing loss of electrical power, including station blackout (SBO). The SBO scenario involves a loss of offsite power, failure of the redundant emergency diesel generators, failure of alternate current (AC) power restoration and the eventual degradation of the reactor coolant pump (RCP) seals resulting in a long term loss of coolant. Using PCTRAN/CPR1000, this study analyses the station blackout on a Chinese CPR1000 which is the most representative type reactor in terms of number of reactors, operating period, power capacity and geological distance from Korean Peninsula. Both the physical effects of the accidents as well as the releases of radioisotopes are calculated and discussed. Station blackout simulation was conducted in this study. The resulting effects seen are consistent with other stress test station blackout tests used utilizing licensed simulation codes. An exact comparison is however not possible as the plants on which the simulations was done vary greatly and the limitations of availability to Chinese FSAR. PCTRAN/CPR1000 is an extremely useful simulation package that provides engineers and scientists very accurate feedback to how a nuclear power plant would react as a whole under various plant conditions. It is able to do this extremely fast as well. As a training tool PCTRAN/CPR1000 provides hands-on experience with many of the primary plant operations and develops an intuitive understanding of the plant.

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

    Directory of Open Access Journals (Sweden)

    Andrej Prošek

    2013-01-01

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

  10. Severe accident insights report

    International Nuclear Information System (INIS)

    Pratt, W.T.

    1988-04-01

    This report describes the conditions and events that nuclear power plant personnel may encounter during the latter stages of a severe core damage accident and what the consequences might be of actions they may take during these latter stages. The report also describes what can be expected of the performance of the key barriers to fission product release (primarily containment systems), what decisions the operating staff may face during the course of a severe accident, and what could result from these decisions based on our current state of knowledge of severe accident phenomena. 9 refs

  11. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1987-01-01

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

  12. Management of severe accidents

    International Nuclear Information System (INIS)

    Jankowski, M.W.

    1988-01-01

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

  13. Analysis of station blackout accidents for the Bellefonte pressurized water reactor

    International Nuclear Information System (INIS)

    Gasser, R.D.; Bieniarz, P.P.; Tills, J.L.

    1986-09-01

    An analysis has been performed for the Bellefonte PWR Unit 1 to determine the containment loading and the radiological releases into the environment from a station blackout accident. A number of issues have been addressed in this analysis which include the effects of direct heating on containment loading, and the effects of fission product heating and natural convection on releases from the primary system. The results indicate that direct heating which involves more than about 50% of the core can fail the Bellefonte containment, but natural convection in the RCS may lead to overheating and failure of the primary system piping before core slump, thus, eliminating or mitigating direct heating. Releases from the primary system are significantly increased before vessel breach due to natural circulation and after vessel breach due to reevolution of retained fission products by fission product heating of RCS structures

  14. Test study on safety features of station blackout accident for nuclear main pump

    International Nuclear Information System (INIS)

    Liu Xiajie; Wang Dezhong; Zhang Jige; Liu Junsheng; Yang Zhe

    2009-01-01

    The theoretical and experimental studies of reactor coolant pump accidents encountered nation-wide and world-wide were described. To investigate the transient hydrodynamic performance of reactor coolant pump (RCP) during the period of rotational inertia in the station blackout accident, some theoretical and experimental studies were carried out, and the analysis of the test results was presented. The experiment parameters, conditions and test methods were introduced. The flow-rate, rotate speed and vibrations were analyzed emphatically. The quadruplicate polynomial curve equation was used to simulate the flow-rate,rotate speed along with time. The test results indicate that the flow-rate and rotator speed decrease rapidly at the very beginning of cut power and the test results accord with the regulation of safety standard. The vibrant displacement of bearing seat is intensified at the moment of lose power, but after a certain period rotor shaft libration changes. The test and analysis results help to understand the hydrodynamic performance of nuclear primary pump under lost of power accident, and provide the basic reference for safety evaluation. (authors)

  15. RCS natural circulation in a PWR station blackout accident--an application of NRC mechanistic codes

    International Nuclear Information System (INIS)

    Han, J.T.

    1987-01-01

    This paper discusses the phenomenon of reactor coolant system (RCS) natural circulation in a PWR station blackout accident with the loss of all AC power and auxiliary feedwater (the TMLB' accident). Existing and future studies performed for the industry and the Nuclear Regulatory Commission (NRC) are summarized in the paper. During the core uncovery and core melt period of the high-pressure TMLB' accident, multi-dimensional natural circulation of gas flow (steam and other gas such as hydrogen and fission products) is likely to exist in the uncovered core and the upper plenum above. Meanwhile, counter-current gas flow may also exist in the hot leg piping except during the opening of a power-operated relief valve (PORV) or safety relief valve (SRV) on the pressurizer. As a result, some of the core decay heat is transferred to the upper plenum structures and ex-vessel piping and components, and the RCS pressure boundary may be heated to high temperature to challenge structural integrity

  16. Evaluation of EBR-II driver-fuel elements following an unprotected station blackout accident

    International Nuclear Information System (INIS)

    Chang, L.K.; Bottcher, J.H.

    1986-01-01

    One of the current design objectives for a liquid metal reactor (LMR) is the inherent shutdown-cooling capability of the reactor, such that the reactor itself can safely reduce power following a total loss of pump power without activating the reactor shutdown system (RSS). Following a loss-of-flow (LOF) accident and a failure of RSS, in EBR-II, reactor core damage and plant restartability is of considerable interest. In the LOF event, high temperature in the reactor causes negative reactivity feedback that reduces reactor power. After an accident, reactor fuel performance is one of the factors used to assess the restartability of the plant. A thermal-hydraulic-neutronic analysis was performed to determine the response of the plant and the temperature of individual subassemblies. These temperatures were then used to assess the damage to driver fuel elements caused by the station blackout accident. The maximum depth of cladding wastage from molten eutectic at temperatures >715 0 C was found to be 0.0053 mm for the hottest subassembly; this value is considerably less than the 0.28 mm cladding thickness. 12 refs

  17. 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.

  18. Development and qualification of a thermal-hydraulic nodalization for modeling station blackout accident in PSB-VVER test facility

    International Nuclear Information System (INIS)

    Saghafi, Mahdi; Ghofrani, Mohammad Bagher; D’Auria, Francesco

    2016-01-01

    Highlights: • A thermal-hydraulic nodalization for PSB-VVER test facility has been developed. • Station blackout accident is modeled with the developed nodalization in MELCOR code. • The developed nodalization is qualified at both steady state and transient levels. • MELCOR predictions are qualitatively and quantitatively in acceptable range. • Fast Fourier Transform Base Method is used to quantify accuracy of code predictions. - Abstract: This paper deals with the development of a qualified thermal-hydraulic nodalization for modeling Station Black-Out (SBO) accident in PSB-VVER Integral Test Facility (ITF). This study has been performed in the framework of a research project, aiming to develop an appropriate accident management support tool for Bushehr nuclear power plant. In this regard, a nodalization has been developed for thermal-hydraulic modeling of the PSB-VVER ITF by MELCOR integrated code. The nodalization is qualitatively and quantitatively qualified at both steady-state and transient levels. The accuracy of the MELCOR predictions is quantified in the transient level using the Fast Fourier Transform Base Method (FFTBM). FFTBM provides an integral representation for quantification of the code accuracy in the frequency domain. It was observed that MELCOR predictions are qualitatively and quantitatively in the acceptable range. In addition, the influence of different nodalizations on MELCOR predictions was evaluated and quantified using FFTBM by developing 8 sensitivity cases with different numbers of control volumes and heat structures in the core region and steam generator U-tubes. The most appropriate case, which provided results with minimum deviations from the experimental data, was then considered as the qualified nodalization for analysis of SBO accident in the PSB-VVER ITF. This qualified nodalization can be used for modeling of VVER-1000 nuclear power plants when performing SBO accident analysis by MELCOR code.

  19. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

    Frid, W.

    1991-08-01

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  1. Extended station blackout analyses of an APR1400 with MARS-KS

    OpenAIRE

    Kim Woongbae; Jang Hyungwook; Oh Seungjong; Lee Sangyong

    2016-01-01

    The Fukushima Daiichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electric energy required for essential systems during a station blackout is provided from emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been exte...

  2. 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.

  3. Severe accident management. Prevention and Mitigation

    International Nuclear Information System (INIS)

    1992-01-01

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

  4. The management of severe accidents

    International Nuclear Information System (INIS)

    Pelce, J.; Brignon, P.

    1987-01-01

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

  5. MELCOR analyses of severe accident scenarios in Oconee, a B ampersand W PWR plant

    International Nuclear Information System (INIS)

    Madni, I.K.; Nimnual, S.; Foulds, R.

    1993-01-01

    This paper presents the results and insights gained from MELCOR analyses of two severe accident scenarios, a Loss of Coolant Accident (LOCA) and a Station Blackout (TMLB) in Oconee, a Babcock ampersand Wilcox (B ampersand W) designed PWR with a large dry containment, and comparisons with Source Term Code Package (STCP) calculations of the same sequences. Results include predicted timing of key events, thermal-hydraulic response in the reactor coolant system and containment, and environmental releases of fission products. The paper also explores the impact of varying concrete type, vessel failure temperature, and break location on the accident progression, containment pressurization, and environmental releases of radionuclides

  6. Blackouts and natural risks

    Science.gov (United States)

    Danihelka, P.; Paldusová, E.; Dobeš, P.

    2009-04-01

    "Blackout" has become the common definition for the situation when electricity supply and demand are not balanced and security of supply fails. These failures have many impacts besides the lights going out, but this term is used commonly. Blackouts have drastic impacts for the society on whole and its citizens and some of them can influence big areas and last for long period, so the consequences are catastrophic. Even if at the European scale, the large extend blackouts are supposed to be exceptional, real frequency is relatively high, approximately once per two years. According to statistics, blackouts are often caused by natural causes, especially lightning. An example of lightning caused blackout is New York blackout 1977, leading to the stand-by of nuclear power plant Indian Point and with overall cost more than 300 mil. USD. There is a clear a distinction between those blackouts caused by nature and those that were caused by other faults. Usually, the nature-caused disturbances as Canada 1988, Sweden 2005 and France 1999, stay inside one country. However, their duration can extend to several weeks, and thus the costs of the interruptions and social impacts are high. Blackouts of only technologic and/or anthropogenic origin are frequently shorter, but may concern more end-users, when cascading from one country to another. Lightning is not the only natural event causing blackouts. Eighteen various case studies of blackout caused by natural events different then lightning were studied and following natural phenomenon found as a root causes: 1x forest fire, 1x snow calamity, 1x ice storm, 1x landslide, 1x high temperature, 1x geomagnetic storm, 2x earthquake, 2x inundation, 2x contact of line with trees, 6x storm (wind, hurricane…). We can conclude, that natural event are frequent cause of blackout of medium or large extend and this phenomena should be studied more in details. This contribution was supported by Ministry of Environment of the Czech Republic.

  7. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  8. 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

  9. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  10. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  11. 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

  12. An uncertainty analysis of the hydrogen source term for a station blackout accident in Sequoyah using MELCOR 1.8.5

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Bixler, Nathan E.; Wagner, Kenneth Charles.

    2014-03-01

    A methodology for using the MELCOR code with the Latin Hypercube Sampling method was developed to estimate uncertainty in various predicted quantities such as hydrogen generation or release of fission products under severe accident conditions. In this case, the emphasis was on estimating the range of hydrogen sources in station blackout conditions in the Sequoyah Ice Condenser plant, taking into account uncertainties in the modeled physics known to affect hydrogen generation. The method uses user-specified likelihood distributions for uncertain model parameters, which may include uncertainties of a stochastic nature, to produce a collection of code calculations, or realizations, characterizing the range of possible outcomes. Forty MELCOR code realizations of Sequoyah were conducted that included 10 uncertain parameters, producing a range of in-vessel hydrogen quantities. The range of total hydrogen produced was approximately 583kg 131kg. Sensitivity analyses revealed expected trends with respected to the parameters of greatest importance, however, considerable scatter in results when plotted against any of the uncertain parameters was observed, with no parameter manifesting dominant effects on hydrogen generation. It is concluded that, with respect to the physics parameters investigated, in order to further reduce predicted hydrogen uncertainty, it would be necessary to reduce all physics parameter uncertainties similarly, bearing in mind that some parameters are inherently uncertain within a range. It is suspected that some residual uncertainty associated with modeling complex, coupled and synergistic phenomena, is an inherent aspect of complex systems and cannot be reduced to point value estimates. The probabilistic analyses such as the one demonstrated in this work are important to properly characterize response of complex systems such as severe accident progression in nuclear power plants.

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

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2014-01-01

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

  14. Severe accident research in France

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  15. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

    1992-12-01

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

  16. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

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

    2012-01-01

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

  17. 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......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

  18. 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)

  19. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  20. NPP Krsko Severe Accident Management Guidelines Upgrade

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  1. HTR-10 severe accident management

    International Nuclear Information System (INIS)

    Xu Yuanhui; Sun Yuliang

    1997-01-01

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

  2. Alcohol-Induced Blackout

    Directory of Open Access Journals (Sweden)

    Dai Jin Kim

    2009-11-01

    Full Text Available For a long time, alcohol was thought to exert a general depressant effect on the central nervous system (CNS. However, currently the consensus is that specific regions of the brain are selectively vulnerable to the acute effects of alcohol. An alcohol-induced blackout is the classic example; the subject is temporarily unable to form new long-term memories while relatively maintaining other skills such as talking or even driving. A recent study showed that alcohol can cause retrograde memory impairment, that is, blackouts due to retrieval impairments as well as those due to deficits in encoding. Alcoholic blackouts may be complete (en bloc or partial (fragmentary depending on severity of memory impairment. In fragmentary blackouts, cueing often aids recall. Memory impairment during acute intoxication involves dysfunction of episodic memory, a type of memory encoded with spatial and social context. Recent studies have shown that there are multiple memory systems supported by discrete brain regions, and the acute effects of alcohol on learning and memory may result from alteration of the hippocampus and related structures on a cellular level. A rapid increase in blood alcohol concentration (BAC is most consistently associated with the likelihood of a blackout. However, not all subjects experience blackouts, implying that genetic factors play a role in determining CNS vulnerability to the effects of alcohol. This factor may predispose an individual to alcoholism, as altered memory function during intoxication may affect an individual‟s alcohol expectancy; one may perceive positive aspects of intoxication while unintentionally ignoring the negative aspects. Extensive research on memory and learning as well as findings related to the acute effects of alcohol on the brain may elucidate the mechanisms and impact associated with the alcohol- induced blackout.

  3. Use of simulators in severe accident management

    International Nuclear Information System (INIS)

    Evans, R.C.

    1994-01-01

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

  4. N Reactor severe accident chemistry

    International Nuclear Information System (INIS)

    Owczarski, P.C.

    1988-01-01

    N Reactor at Hanford has a number of features that are unique compared to commercial LWRs. These features can affect the outcome of postulated core-damage accidents. The massive metallic uranium fuel at low burn-up can delay core melting and, along with reducing conditions, keep fission product release and aerosol particle masses low. The horizontal pressure tube arrangement in the massive graphite moderator can keep damaged fuel from contacting large amounts of water, thus limiting the amount of hydrogen produced. Large surface areas in the primary piping and the fog sprays can remove airborne aerosol particles and vapors to the point where noble gases can become the dominant dose contributors. The fog spray systems can wash out up to 98% of the released particulate and vapor fission products, creating a unique liquid effluent

  5. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

    Clauss, D.B.

    1989-11-01

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

  6. Chemical considerations in severe accident analysis

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  7. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

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

    1994-01-01

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

  8. Instrumentation and severe accident plant status interpretation

    International Nuclear Information System (INIS)

    Chao, J.; Machiels, A.J.; Oehlberg, R.N.; Negin, C.A.; James, R.

    1992-01-01

    EPRI is conducting a project related to instrumentation and severe accident plant status interpretation. The project will recognize the facts that (i) instrument responses during severe accidents do not need to be as accurate as during normal operation, and (ii) not all instrument loops will see a severe environment. In particular, the proposed work is to provide technology to get the most information from the existing instrumentation under severe accident conditions by developing (1) calculational aids to determine actual plant parameters based on severe-accident-affected instrument readings, and (2) means to utilize indications from operational instruments to infer parameters values for failed instruments, or where no instrument may exist. Specific deliverables for this project are (i) an instrumentation data base that will include both instrumentation failures and successes under severe conditions, and contain instrument performance information from both nuclear and non-nuclear industry situations; (ii) methods to assess the validity of instrument signals and estimate the performance of individual instrument loops; and (iii) calculational aids to estimate and interpret instrument readings under severe accident conditions, including the ability to extrapolate readings from functioning instruments to locations where instruments have failed. (orig.)

  9. 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

  10. Overview of severe accident research at JAERI

    International Nuclear Information System (INIS)

    Sugimoto, Jun

    1999-01-01

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

  11. Application of FFTBM to severe accidents

    International Nuclear Information System (INIS)

    Prosek, A.; Leskovar, M.

    2005-01-01

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

  12. Severe accident testing of a personnel airlock

    International Nuclear Information System (INIS)

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

    1988-01-01

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

  13. Core loss during a severe accident (COLOSS)

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-08-15

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

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

    International Nuclear Information System (INIS)

    2004-01-01

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

  17. 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.

  18. APR1400 severe accident mitigation design

    Energy Technology Data Exchange (ETDEWEB)

    Jae Young, Lim; Jae Youb, Byun [Shin-Kori 3 and 4 NPP Project, Korea Power Engineering Company, Yongin-si, Gyeonggi-do (Korea, Republic of)

    2007-07-01

    APR1400, a Korean evolutionary advanced LWR, has been developed to meet the quantitative safety goals of mean core damage frequency to be less than one in one hundred thousand reactor years (10{sup -5}/y) and the expected overall mean frequency of occurrence of offsite doses in excess of 0.01 Sv within 24 hours at the site boundary to be less than one per million reactor years (10{sup -6}/y). In order to meet these quantitative goals, defense in depth, a long standing fundamental principle of reactor safety, was applied to ensure plant safety and to provide the balanced design between prevention and mitigation. And various advanced design features were reviewed to improve plant safety in the viewpoint of prevention and mitigation of design basis accident and severe accident. In this paper, 5 issues concerning severe accident mitigation features of the APR1400 are reviewed: 1) hydrogen control, 2) high pressure melt ejection and direct containment heating, 3) steam explosion, 4) molten corium concrete interaction, and 5) equipment survivability. It is shown that the APR1400 has been designed to withstand severe accidents.

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

    International Nuclear Information System (INIS)

    Basic, I.; Kocnar, R.

    1996-01-01

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

  20. Occupational Radiation Protection in Severe Accident Management

    International Nuclear Information System (INIS)

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  2. Some considerations on severe accident countermeasures

    International Nuclear Information System (INIS)

    Tominaga, Kenji; Ohtani, Masanori

    2015-01-01

    The Severe Accident (SA) leading to core melt had not been considered to occur in Japanese Nuclear Power Plants (NPPs) because various kinds of countermeasures were adopted for the first through third layers of Defense in Depth (DiD). The direct cause of Fukushima Daiichi accident was considered to be the loss of all AC powers caused by an unexpected huge tsunami, but the insufficient measures for the fourth and fifth layer of DiD was pointed out as a lessons learned of Fukushima Daiichi accident. From this viewpoint, foreign NPP investigations have been performed by JANSI in order to gather and analyze information how foreign NPPs are taking SA countermeasures. Before Fukushima Daiichi accident, SA countermeasures adopted in Japanese NPPs were far behind from the advanced foreign NPPs, but nowadays we think that SA countermeasures in Japanese NPPs have caught up and have reached almost the same level as advanced foreign NPPs. Moreover, JANSI is going to study additional SA countermeasures as voluntary basis for the safety improvement of Japanese NPPs based on SRS-46 and foreign NPPs investigations. (author)

  3. Nuclear power plant Severe Accident Research Plan

    International Nuclear Information System (INIS)

    Larkins, J.T.; Cunningham, M.A.

    1983-01-01

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

  4. Development of Severe Accident Containment Analysis Package

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  5. 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.

  6. Sarnet lecture notes on nuclear reactor severe accident phenomenology

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  7. Development of a totally integrated severe accident training system

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  8. 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.

  9. Study Of Severe Accident Phenomena In Nuclear Power Plant

    International Nuclear Information System (INIS)

    Sugiyanto; Antariksawan; Anhar, R.; Arifal

    2001-01-01

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

  10. Strategy Evaluation for Cavity Flooding during an ESBO Initiated Severe Accident

    Directory of Open Access Journals (Sweden)

    Nan Jiang

    2018-01-01

    Full Text Available Intentional depressurization and cavity flooding are two important measures in current severe accident management guidelines (SAMGs. An extreme scenario of an extended station blackout (ESBO, when electric power is unavailable for more than 24 hours, actually occurred in the Fukushima Daiichi accident and attracted lots of attention. In an ESBO, the containment spray cannot be activated for condensation, and, thus, cavity flooding will generate a large amount of steam, which, ironically, overpressurizes the containment to failure before the reactor vessel is melted through. Therefore, consideration of these conflicting issues and the ways in which plants operate is crucial for strengthening the strategies outlined in SAMGs. In this paper, the effects of intentional depressurization and cavity flooding in an ESBO for a representative 900 MW second-generation pressurized water reactor (PWR are simulated with MAAP4 code. Diverse scenarios with different starting times of depressurization and water injection are also compared to summarize the positive and negative impacts for accident mitigation. The phenomena associated with creep ruptures, hydrogen combustion, corium stratification, and cavity boiling are also analyzed in detail to strengthen our understanding of severe accident mechanisms. The results point out the facility limitations of second-generation PWRs which can improve existing SAMGs.

  11. A new perspective on severe nuclear accidents

    International Nuclear Information System (INIS)

    Lee, Jaiki

    2012-01-01

    The reactions of the public in Korea to the nuclear accident at the Fukushima Daiichi plants in Japan, particularly over-reactions, are reviewed, with the conclusion that significant radioactive contamination of a small country could lead to a severe national crisis. The most important factor is the socio-economic damage caused by stigma, which in turn is caused by a misunderstanding of the radiation risk. Given that nuclear power is an important choice in the face of the threat of climate change, the public's perceptions need to be changed at any cost, not only in those countries operating nuclear power plants but globally as well. (note)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-02-15

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

  13. Severe Accidents in the Energy Sector

    International Nuclear Information System (INIS)

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

    1998-11-01

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

  14. 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

  15. Pressurized-water-reactor station blackout

    International Nuclear Information System (INIS)

    Dobbe, C.A.

    1983-01-01

    The purpose of the Severe Accident Sequence Analysis (SASA) Program was to investigate accident scenarios beyond the design basis. The primary objective of SASA was to analyze nuclear plant transients that could lead to partial or total core melt and evaluate potential mitigating actions. The following summarizes the pressurized water reactor (PWR) SASA effort at the Idaho National Engineering Laboratory (INEL). The INEL is presently evaluating Unresolved Safety Issue A-44 - Station Blackout from initiation of the transient to core uncovery. The balance of the analysis from core uncovery until fission product release is being performed at Sandia National Laboratory (SNL). The current analyses involve the Bellefonte Nuclear Steam Supply System (NSSS), a Babcock and Wilcox (B and W) 205 Fuel Assembly (205-FA) raised loop design to be operated by the Tennessee Valley Authority

  16. Porosity effects during a severe accident

    International Nuclear Information System (INIS)

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

    2015-09-01

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

  17. 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)

  18. Extended Station Blackout Analyses of an APR1400 with MARS-KS

    International Nuclear Information System (INIS)

    Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong

    2016-01-01

    The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage

  19. Extended Station Blackout Analyses of an APR1400 with MARS-KS

    Energy Technology Data Exchange (ETDEWEB)

    Kim, WoongBae; Jang, HyungWook; Oh, Seungjong; Lee, Sangyong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

    2016-10-15

    The Fukushima Dai-ichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electricity required for essential systems during a station blackout is provided from the emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6 and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating of extended station blackout scenarios, the best estimate MARS-KS was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study was performed on reactor coolant pump seal leakage.

  20. Extended station blackout analyses of an APR1400 with MARS-KS

    Directory of Open Access Journals (Sweden)

    Kim Woongbae

    2016-01-01

    Full Text Available The Fukushima Daiichi nuclear power plant accident shows that natural disasters such as earthquakes and the subsequent tsunamis can cause station blackout for several days. The electric energy required for essential systems during a station blackout is provided from emergency backup batteries installed at the nuclear power plant. In South Korea, in the event of an extended station blackout, the life of these emergency backup batteries has recently been extended from 8 hours to 24 hours at Shin-Kori 5, 6, and APR1400 for design certification. For a battery life of 24 hours, available safety means system, equipment and procedures are studied and analyzed in their ability to cope with an extended station blackout. A sensitivity study of reactor coolant pump seal leakage is performed to verify how different seal leakages could affect the system. For simulating extended station blackout scenarios, the best estimate MARS-KS computer code was used. In this paper, an APR1400 RELAP5 input deck was developed for station blackout scenario to analyze operation strategy by manually depressurizing the reactor coolant system through the steam generator's secondary side. Additionally, a sensitivity study on reactor coolant pump seal leakage was carried out.

  1. SARNET: Severe accident research network of excellence

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  2. 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

  3. Aerosol transport in severe reactor accidents

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  4. Severe accident management guidelines (SAMGs) for German NPPs; Severe Accident Management Guidelines (SAMGS) fuer deutsche KKW

    Energy Technology Data Exchange (ETDEWEB)

    Bauer, Martin; Tietsch, Wolfgang [Westinghouse Electric Germany GmbH, Mannheim (Germany)

    2011-07-01

    The US NRC declared in consequence of the Three Mile Island accident the severe accident management to be an unresolved safety issue. In the following years worldwide the development of generic management guidelines was started aimed to implement preventive measures to prevent core damage and mitigation measures to reduce the accident consequences into the emergency manuals of nuclear power plants. The PWROG (pressurized water reactor owners group) SAMGs were developed by Westinghouse that are meanwhile implemented in US NPPs but also in Europe, incl. reactors with VVER, KWU or Framatome design. The authors describe the concept of the emergency manuals implemented in German NPPs and the differences to the SAMG concept. SAMGs include a complete strategy to minimize the consequences of severe accidents, independent of the plant status, including the risk of component failures. Measures to bring the meltdown to an end have not priority. The implementation of SAMGs into the German emergency manuals needs clear criteria for the transition from preventive measures (aimed to stabilize the reactor) to mitigation measures (minimization of fission product release). Several examples are discussed.

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

  6. Study of the causes of pedestrian accidents by severity | Kouabenan ...

    African Journals Online (AJOL)

    In the conclusion, we point out the utility of an approach that combines several methods of accident analysis and we consider some ways to improve the use of accident reports for prevention purposes. Keywords: accident report, causal attribution, causality tree. Trois méthodes différentes sont appliquées à l'analyse de 55 ...

  7. Tchernobyl: a severe accident and its image

    International Nuclear Information System (INIS)

    Strazzulla, J.

    1996-01-01

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

  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

    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)

  10. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

    1984-03-01

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

  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. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    2008-06-01

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

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

    International Nuclear Information System (INIS)

    Sandoval V, S.

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-03-15

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

  15. 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.

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

    International Nuclear Information System (INIS)

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

    1993-09-01

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

  17. Reactor Cavity in Case of Station Blackout in RBMK-1500

    Directory of Open Access Journals (Sweden)

    Algirdas Kaliatka

    2007-01-01

    Full Text Available Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA to a severe accident are discussed.

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

    African Journals Online (AJOL)

    Weighted Severity Index (WSI) was created based on these factors/drivers. Also, Density-based Clustering for Traffic Accident Risk (DBCTAR) was carried out to assist in ascertaining the distribution of Black Spots Severity (BSS). Results obtained include: shortestpath analysis, service area analysis, accident spot severity ...

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  20. Evaluation of a cavity flooding strategy for the prevention of reactor vessel failure in a severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Park, Rae Joon; Je, Moo Sung; Park, Chang Kyoo [Korea Atomic Energy Research Institute, TaeJon (Korea, Republic of)

    1994-10-01

    As a part of the evaluation of accident management strategies for severe accident prevention or mitigation in a station blackout scenario for YGN 3 and 4, an external vessel cooling strategy for the prevention of reactor vessel failure has been estimated using the MAAP4 computer code. The sensitivity studies have been performed such as actuating timings and the number of spray pumps used. To explore external vessel cooling strategies, containment spray pumps were actuated by varying time spanning core uncovery, core melting and relocation of molten core material. It was shown that flooding of the reactor cavity using the containment spray system may prevent reactor vessel failure but may not prevent the failure of the relocation of molten core material during the station blackout sequence of YGN 3 and 4. Reactor vessel failure can be prevented by external vessel cooling using condensed water from the operation of two containment spray pumps at the time of core melting and using water from the operation of one containment spray pumps at the time of core melting and using water from the operation of one containment spray pump at the time of core uncovery. (Author) 46 refs., 26 figs., 5 tabs.

  1. Analysis of two different types of hydrogen combustion during severe accidents in a typical pressurized water reactor

    International Nuclear Information System (INIS)

    Ko Yuchih; Lee Min

    2005-01-01

    Hydrogen combustion is an important phenomenon that may occur during severe accidents of Nuclear Power Plants (NPPs). Depending on the specific plant design, the initiating events, and mitigation actions executed, hydrogen combustion may have distinct characteristics and may damage the plant in various degrees. The worst scenario will be the catastrophic failure of containment. In this study two specific types of hydrogen combustion are analyzed to evaluate their impact on the containment integrity. In this paper, Station Blackout (SBO) and Loss of Coolant Accidents (LOCAs) sequences are analyzed using MAAP4 (Modular Accident Analysis Program) code. The former sequence is used to represent hydrogen combustion phenomenon under the condition that the reactor pressure vessel (RPV) breaches at high pressure and the latter sequence represents the phenomenon that RPV fails at low pressure. Two types of hydrogen combustion are observed in the simulation. The Type I hydrogen combustion represents global and instantaneous hydrogen combustion. Large pressure spike is created during the combustion and represents a threat to containment integrity. Type II hydrogen combustion is localized burn and burn continuously over a time period. There is hardly any impact of this type hydrogen burn on the containment pressurization rate. Both types of hydrogen combustion can occur in the severe accidents without any human intervention. From the accident mitigation point of view, operators should try to bring the containment into conditions that favor the Type II hydrogen combustion. (authors)

  2. Severe Accident Management Strategy for EU-APR1400

    International Nuclear Information System (INIS)

    Hwang, Do Hyun; Kim, Yong Soo; Yoon, Sun Hong

    2013-01-01

    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

  3. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

    1989-04-01

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

  4. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-01-01

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

  5. The Effect of Containment Filtered Venting System on the Severe Accident Management Strategies of the CANDU6 Plant

    Energy Technology Data Exchange (ETDEWEB)

    Jin, Youngho [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    In March, 2011, Fukushima daichi nuclear power plants experienced a long term station blackout and severe core damages and released a large amount of radioactive materials outside of the plants. After this accident Nuclear Safety and Security Commission (NSSC) decided to install a filtered containment venting system (CFVS) at all the operating nuclear power plants in Korean. To comply with NSSC's request, Wolsong Unit 1 has installed a CFVS. Current severe accident management guidance, which does not consider a CFVS has 6 severe accident management strategies for CANDU6 plant. These strategies are inject in to the primary heat transport system (PHTS), inject in to the calandria, inject into the calandria vault, reduce fission product releases, control containment conditions, reduce containment hydrogen. The CFVS is designed to open and to close isolation valves by an operator. An operator opens the CFVS isolation valve when the containment pressure exceeds the design pressure (124 kPa(g)) and closes isolation valves when the containment pressure decreases below 50 kPa(g). The operation of the CFVS not only influences the current strategies (adds a means of controlling containment conditions) but also requires the new strategies. This paper discusses the necessity of the new strategies, such as the prevention of containment vacuum and the injection into the containment. The necessity of the additional severe accident management strategies for CANDU6 plants which installed a CFVS is evaluated. The operation of a CFVS affects the water inventory in the basement also, but not significantly. The SBO accident requires the water injection into the containment at least 4 days after an accident initiation if a passive spray system fails. If a spray system operates, then the injection into the containment is required more than 10 days after an accident initiation even though a CFVS operates.

  6. Using MARS to assist in managing a severe accident

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  7. Management of severe pelvic injury following road traffic accident in ...

    African Journals Online (AJOL)

    A 34 year old woman involved in road traffic accident with severe anterior and posterior pelvic fractures with associated soft tissue injury was referred from Wa Regional Hospital 18 hours after the accident to Tania Specialist Hospital in Tamale. Emergency resuscitative measures such as catheterization and management of ...

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

    International Nuclear Information System (INIS)

    Jeong, Kwangsub; Ha, Jaejoo

    2000-01-01

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

  9. The philosophy of severe accident management in the US

    International Nuclear Information System (INIS)

    Baratta, A.J.

    1990-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Lombardo, N J; Lanning, D D; Panisko, F E [Pacific Northwest Lab., Richland, WA (United States)

    1992-12-01

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

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

    International Nuclear Information System (INIS)

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

    1992-12-01

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

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

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  13. 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).

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

    Directory of Open Access Journals (Sweden)

    JIN HO SONG

    2014-04-01

    Adequacy of current safety goals are also discussed in view of the socio-economic impact of the Fukushima accident. As a conclusion, it is suggested that an investigation on a coherent integrated safety principle for the severe accident and development of innovative mitigation features is necessary for robust and resilient nuclear power system.

  15. Insights on severe accident chemistry from TMI-2

    International Nuclear Information System (INIS)

    Hobbins, R.R.; Cronenberg, A.W.; Langer, S.; Owen, D.E.; Akers, D.W.

    1986-01-01

    Chemical analyses are being carried out on materials removed from the damaged reactor at TMI-2. Characteristics of TMI-2 fuel, control, fission product and structural materials based on these analyses are presented. Emphasis is placed on chemistry within the pressure vessel although descriptions of, and postulated mechanisms for, materials transported from the vessel to the reactor building are also discussed. Indications of the oxygen potential in the reactor pressure vessel during the high temperature phase of the accident are of particular significance for the analysis of damage progression and fission product behavior during severe accidents. The results of thermodynamic and kinetic calculations for chemical species present during the high temperature portion of the accident (during core uncovery) are presented. Insights on chemistry of significance for severe accident analysis which follow from the evaluation of the TMI-2 accident are discussed. 38 refs., 9 figs., 2 tabs

  16. Conditions for oxygen-deficient combustion during accidents with severe core concrete thermal attack

    International Nuclear Information System (INIS)

    Luangdilok, W.; Elicson, G.T.; Berger, W.E. Jr.

    1993-01-01

    This paper addresses the interactions between MCCI (molten core-concrete interactions)-induced offgas releases, mostly the combustible gases, natural circulation between the cavity and the lower containment based on recent research developments in the area of mixed convection flow (Epstein, et al., 1989; Epstein, 1988; Epstein, 1992) between compartments, and their effects on combustion in PWR containments during prolonged severe accidents. Specifically, large dry PWR containments undergoing severe core-concrete attack during station blackouts where the containment atmosphere is expected to be inerted are objects of this analysis. The purpose of this paper, given the conditions that oxygen can be brought to the cavity, is to demonstrate that consumption of most oxygen present in the containment can be achieved in a reasonable time scale assuming that combustion is not subject to flammability limits due to the high cavity temperatures. The conditions for cavity combustion depend on several factors including good gas flowpaths between the cavity and other containment regions, and combustion processes within the cavity with the hot debris acting as the ignition source

  17. Detailed evaluation of melt pool configuration in the lower plenum of the APR1400 reactor vessel during severe accidents

    International Nuclear Information System (INIS)

    Park, Rae-Joon; Kang, Kyoung-Ho; Hong, Seong-Wan; Kim, Hwan-Yeol

    2015-01-01

    Highlights: • Melt pool condition in the lower plenum was determined using SCDAP/RELAP5. • GEMINI analyses were performed to examine the final melt pool composition. • A density evaluation graph was developed for the melt pool layer inversion. • The final melt pool configurations were determined for five accident scenarios. • The thermodynamic results address the possibility of the layer inversion. - Abstract: For a detailed evaluation of the IVR (In-Vessel corium Retention) through the ERVC (External Reactor Vessel Cooling) during a severe accident, the melt pool configuration should be accurately determined in the lower plenum of the reactor vessel. It affects the thermal load to the vessel wall and plays a key role in determining the integrity of the reactor vessel under the IVR-ERVC. SCDAP/RELAP5 and GEMINI analyses have been performed to determine the final corium condition and examine the final melt pool composition at a reactor vessel failure during a severe accident in an APR (Advanced Power Reactor) 1400, respectively. As the representative accident scenarios, five dominant sequences of the TLFW (Total Loss of Feed Water), the SBO (Station BlackOut), the SBLOCA (Small Break Loss of Coolant Accident) without SI (Safety Injection), the MBLOCA without SI, and the LBLOCA without SI were selected from the level I PSA (Probabilistic Safe Assessment) results. A density evaluation graph was developed for the precise examination of the melt pool layer inversion. The final melt pool configurations at the reactor vessel failure have been determined for five dominant accident scenarios of the APR1400 using the GEMINI results and the density evaluation graph. The thermodynamic analysis results in three sequences of the APR1400 accident address the possibility of a melt pool layer inversion in the lower plenum of the reactor vessel. The layer inversion led to corium pool stratification with a heavy metallic layer below the oxidic pool, which leads to a three

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  19. Modelling of blackout sequence at Atucha-1 using the MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    This paper presents the modelling of a complete blackout at the Atucha-1 NPP as preliminary phase for a Level II safety probabilistic analysis. The MARCH3 code of the STCP (Source Term Code Package) is used, based on a plant model made in accordance with particularities of the plant design. The analysis covers all the severe accident phases. The results allow to view the time sequence of the events, and provide the basis for source term studies. (author). 6 refs., 2 figs

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

    International Nuclear Information System (INIS)

    Marino, G.P.

    1986-04-01

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

  1. Simulator drills for the management of severe accidents

    International Nuclear Information System (INIS)

    Hoffmann, E.

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Cizel, F.

    1999-01-01

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

  3. 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.

  4. Severities of transportation accidents involving large packages

    International Nuclear Information System (INIS)

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

    1978-05-01

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

  5. Structural evaluation of electrosleeved tubes under severe accident transients

    International Nuclear Information System (INIS)

    Majumdar, S.

    1999-01-01

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

  6. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

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

    1998-01-01

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

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

    International Nuclear Information System (INIS)

    1989-08-01

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

  8. Parametric and experimentally informed BWR Severe Accident Analysis Utilizing FeCrAl - M3FT-17OR020205041

    Energy Technology Data Exchange (ETDEWEB)

    Ott, Larry J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2017-08-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered as advanced fuel cladding concepts with enhanced accident tolerance. At high temperatures, FeCrAl alloys have slower oxidation kinetics and higher strength compared with zirconium-based alloys. FeCrAl could be used for fuel cladding and spacer or mixing vane grids in light water reactors and/or as channel box material in boiling water reactors (BWRs). There is a need to assess the potential gains afforded by the FeCrAl accident-tolerant-fuel (ATF) concept over the existing zirconium-based materials employed today. To accurately assess the response of FeCrAl alloys under severe accident conditions, a number of FeCrAl properties and characteristics are required. These include thermophysical properties as well as burst characteristics, oxidation kinetics, possible eutectic interactions, and failure temperatures. These properties can vary among different FeCrAl alloys. Oak Ridge National Laboratory has pursued refined values for the oxidation kinetics of the B136Y FeCrAl alloy (Fe-13Cr-6Al wt %). This investigation included oxidation tests with varying heating rates and end-point temperatures in a steam environment. The rate constant for the low-temperature oxidation kinetics was found to be higher than that for the commercial APMT FeCrAl alloy (Fe-21Cr-5Al-3Mo wt %). Compared with APMT, a 5 times higher rate constant best predicted the entire dataset (root mean square deviation). Based on tests following heating rates comparable with those the cladding would experience during a station blackout, the transition to higher oxidation kinetics occurs at approximately 1,500°C. A parametric study varying the low-temperature FeCrAl oxidation kinetics was conducted for a BWR plant using FeCrAl fuel cladding and channel boxes using the MELCOR code. A range of station blackout severe accident scenarios were simulated for a BWR/4 reactor with Mark I containment. Increasing the FeCrAl low-temperature oxidation rate

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  10. 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.

  11. Development of Extended Station Blackout Recovery Guideline for OPR1000 and APR1400

    International Nuclear Information System (INIS)

    Lee, Chang Gyun; Cho, Dong Hyun; Park, Ki Moon; Huh, Jae Young; Lee, Gyu Cheon

    2016-01-01

    Many regulatory requirements and recommendations following the Fukushima accident have been issued to cope with the extended station blackout (SBO) by the NRC, INPO, IAEA, ENSREG, WENRA, etc., and the nuclear safety improvement design features of each country have been enhanced to incorporate the lessons learned from the Fukushima accident. There have been many evaluations to cope with the extended loss of alternating current (AC) power (ELAP) event after the Fukushima accident. PWROG has developed the FLEX support guideline (FSG) that provides the guidance to mitigate the consequences of ELAP event based on the FLEX. The FSG is interfaced with emergency operating guidelines (EOGs) and severe accident management guidelines (SAMGs). However, the FSG developed by PWROG is not compatible with EOGs for both OPR1000 and APR1400 NPPs. Therefore, it is necessary to develop an extended station blackout recovery guideline (ESRG) to cope with an extended SBO event utilizing the newly adopted safety improvement design features against Fukushima accident for OPR1000 and APR1400 NPPs. The ESRG is also performed to satisfy all safety functions and to prevent from entering SAMGs during an extended SBO event. Therefore, this ESRG is entirely appropriate to cope with an extended SBO event by utilizing the newly adopted safety improvement design features following Fukushima accident for OPR1000 and APR1400 NPPs. This guideline will be considered in the establishment of accident management planning in near future.

  12. AP1000R severe accident features and post-Fukushima considerations

    International Nuclear Information System (INIS)

    Scobel, J. H.; Schulz, T. L.; Williams, M. G.

    2012-01-01

    The AP1000 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)

  13. 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)

  14. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    Silverman, Eugene B.

    2004-01-01

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

  15. Evaluation of severe accident risks and the potential for risk reduction: Grand Gulf, Unit 1. Draft for comment, February 1987

    International Nuclear Information System (INIS)

    Amos, C.N.; Benjamin, A.S.; Kunsman, D.M.; Williams, D.C.; Boyd, G.J.; Lewis, S.R.; Smith, L.N.

    1987-04-01

    The Severe Accident Risk Reduction Program (SARRP) has completed a rebaselining of the risks to the public from a boiling water reactor with a Mark III containment (Grand Gulf, 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 low relative to previous studies; for example, most of the uncertainty range is lower than the point estimate of risk for the Peach Bottom plant in the Reactor Safety Study (RSS). However, certain unresolved issues cause the top of the uncertainty band to appear at a level that is comparable with the RSS point estimate. These issues include the diesel generator failure rate, iodine and cesium revolatilization after vessel breach and the possibility of reactor vessel pedestal failure caused by core debris attack. Some of the postulated safety options appear to be potentially cost effective for the Grand Gulf power plant, particularly when onsite accidents costs are included in the evaluation of benefits. Principally these include procedural modifications and relatively inexpensive hardware additions to insure core cooling in the event of a station blackout. This work supports the Nuclear Regulatory Commission's assessment of severe accidents in NUREG-1150. (author)

  16. Natural hazard impact on the technosphere: "blackouts

    Science.gov (United States)

    Petrova, E. G.

    2012-04-01

    In recent years, natural-technological accidents (NTA) and disasters are increasing in their number and severity all over the world. The term "natural-technological accident (disaster)" applies for an accident (disaster) in the technosphere triggered by any natural process or phenomenon. Their growth is caused, on the one hand, by observed increasing in the frequency and intensity of some natural hazards and hazardous events due to climate change and, on the other hand, by a growing complication of the modern technosphere exposed to natural impacts and advancement of economic activities into the area at natural risk. The most large-scaled natural-technological disaster happened on March 11, 2011 in Japan, as a result of a massive earthquake and tsunami that caused a number of serious technological accidents, including accidents at "Fukushima-1" nuclear power plant, etc. Severe social, ecological and economic consequences of large-scaled NTA make investigation of these events especially important. The most frequent among NTA occurring in Russia are breakdowns in electric power supply systems that lead to so-called "blackouts" (accidental power outages). They are mainly caused by strong winds, snowstorms, deposition of ice, sleet, and snow, rainfalls, floods, and hailstones. Among other triggers earthquakes, hard frost, fierce heat, thunderstorms, landslides, snow avalanches, and debris flows should be mentioned. The great part of transmission facilities in Russia falls on overhead lines that are especially vulnerable to natural impacts. In general, natural triggers are responsible for more than 70 percent of all accidents in power supply systems. They occur more often in Far East, in the Southern and North-Western federal districts, and in some regions of the Central Russia, which are prone to hurricanes, cyclones, snowstorms, and heavy rainfalls accompanying by hailstones, icing, and sleet. A distinctive feature of these events is their synergistic nature, as power

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    Suh, Nam Duk; Park, Jae Hong

    2005-01-01

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

  19. Method to stop severe accident in nuclear reactors

    International Nuclear Information System (INIS)

    Saxena, A.K.; Limaye, S.P.

    2011-01-01

    'Prevent Problems Before they occur' a concept which is easy to understand but difficult to practicise when it applies particularly to nuclear reactors where managing severe accident is an uphill task. Understanding the mechanism of occurrence of split or rupture in clad is still a challenge. Simultaneous occurrence of split or rupture in number of channels easily leads to severe accident. The paper presents a method not to manage severe accidents but to prevent the same to occur in nuclear reactors based on deterministic approach. A computer program is developed with consideration of complex thermal hydraulics and severe accident phenomenon to ascertain the minimum flow rate of coolant above which there are practically no chance of any split or rupture to occur. The results of program are in close agreement with authors own experimental data and the data available by various other researchers and are discussed at length in the paper. An important objective is to pinpoint the location of the likely split or rupture in clad subsequent to LOCA. In-depth study of severe accident is made by considering the clad alone and filled with fuel separately. (author)

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

    International Nuclear Information System (INIS)

    1988-01-01

    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

  1. Design consideration on severe accident for future LWR

    International Nuclear Information System (INIS)

    Omoto, A.

    1998-01-01

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

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

    International Nuclear Information System (INIS)

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

    1984-01-01

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

  3. ESTABLISHMENT OF A SEVERE ACCIDENT MITIGATION STRATEGY FOR AN SBO AT WOLSONG UNIT 1 NUCLEAR POWER PLANT

    Directory of Open Access Journals (Sweden)

    SUNGMIN KIM

    2013-08-01

    Full Text Available During a station blackout (SBO, the initiating event is a loss of Class IV and Class III power, causing the loss of the pumps, used in systems such as the primary heat transporting system (PHTS, moderator cooling, shield cooling, steam generator feed water, and re-circulating cooling water. The reference case of the SBO case does not credit any of these active heat sinks, but only relies on the passive heat sinks, particularly the initial water inventories of the PHTS, moderator, steam generator secondary side, end shields, and reactor vault. The reference analysis is followed by a series of sensitivity cases assuming certain system availabilities, in order to assess their mitigating effects. This paper also establishes the strategies to mitigate SBO accidents. Current studies and strategies use the computer code of the Integrated Severe Accident Analysis Code (ISAAC for Wolsong plants. The analysis results demonstrate that appropriate strategies to mitigate SBO accidents are established and, in addition, the symptoms of the SBO processes are understood.

  4. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

    1987-01-01

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

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

    Science.gov (United States)

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

    2018-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Doizi D.

    2018-01-01

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

  7. Severe accident considerations in Canadian nuclear power reactors

    International Nuclear Information System (INIS)

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

    1990-08-01

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

  8. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-12-01

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

  12. 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

  13. Steam Oxidation Testing in the Severe Accident Test Station

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

    After the March 2011 accident at Fukushima Daiichi, Oak Ridge National Laboratory (ORNL) began conducting high temperature steam oxidation testing of candidate materials for accident tolerant fuel (ATF) cladding in August 2011 [1-11]. The ATF concept is to enhance safety margins in light water reactors (LWR) during severe accident scenarios by identifying materials with 100× slower steam oxidation rates compared to current Zr-based alloys. In 2012, the ORNL laboratory equipment was expanded and made available to the entire ATF community as the Severe Accident Test Station (SATS) [4,12]. Compared to the current UO2/Zr-based alloy fuel system, an ATF alternative would significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident [13-14]. The steam oxidation behavior of candidate materials is a key metric in the evaluation of ATF concepts and also an important input into models [15-17]. However, initial modeling work of FeCrAl cladding has used incomplete information on the physical properties of FeCrAl. Also, the steam oxidation data being collected at 1200°-1700°C is unique as no prior work has considered steam oxidation of alloys at such high temperatures. Also, because many accident scenarios include steadily increasing temperatures, the required data are not traditional isothermal exposures but exposures with varying “ramp” rates. In some cases, the steam oxidation behavior has been surprising and difficult to interpret. Thus, more fundamental information continues to be collected. In addition, more work continues to focus on commercially-manufactured tube material. This report summarizes recent work to characterize the behavior of candidate alloys exposed to high temperature steam, evaluate steam oxidation behavior in various ramp scenarios and continue to collect integral data on FeCrAl compared to conventional Zr-based cladding.

  14. Overview of SAMPSON code development for LWR severe accident analysis

    International Nuclear Information System (INIS)

    Naitoh, Masanori

    2006-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has developed a severe accident analysis code 'SAMPSON'. SAMPSON's distinguishing features include inter-connected hierarchical modules and mechanistic models covering a wide spectrum of scenarios ranging from normal operation to hypothetical severe accident events. Each module included in the SAMPSON also runs independently for analysis of specific phenomena assigned. The OECD International Standard Problems (ISP-45 and 46) were solved by the SAMPSON for code verifications. The analysis results showed fairly good agreement with the test results. Then, severe accident phenomena in typical PWR and BWR plants were analyzed. The PWR analysis result showed 56 hours as the containment vessel failure timing, which was 9 hours later than one calculated by MELCOR code. The BWR analysis result showed no containment vessel failure during whole accident events, whereas the MELCOR result showed 10.8 hours. These differences were mainly due to consideration of heat release from the containment vessel wall to atmosphere in the SAMPSON code. Another PWR analysis with water injection as an accident management was performed. The analysis result showed that earlier water injection before the time when the fuel surface temperature reached 1,750 K was effective to prevent further core melt. Since fuel surface and fluid temperatures had spatial distribution, a careful consideration shall be required to determine the suitable location for temperature measurement as an index for the pump restart for water injection. The SAMPSON code was applied to the accident analysis of the Hamaoka-1 BWR plant, where the pipe ruptured due to hydrogen detonation. The SAMPSON had initially been developed to run on a parallel computer. Considering remarkable progress of computer hardware performance, as another version of the SAMPSON code, it has recently been modified so as to run on a single processor. The improvements of physical models, numerical

  15. 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.

  16. The development of a severe accident analysis code

    International Nuclear Information System (INIS)

    Kim, Hee Dong; Cho, S. W.; Park, J. H.; Hong, S. W.; Hwang, M. K.; Kim, D. H.; Park, S. Y.; Kim, S. D.; Nho, K. M.

    1997-07-01

    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 in an effect to improve existing models and develop analytical tools for the assessment of severe accidents. For hydrogen control, the analysis of hydrogen concentration in the containment and visualization for the concentration in the cell were performed. The computer code to predict combustion flame characteristic was also developed. the analytical model for the expansion phase of vapor explosion was developed and verified with the experimental results. The corium release fraction model from the cavity with the capture volume was developed and applied to the power plants. Pre-test calculation was performed for molten corium concrete interaction study and the crust formation process, heat transfer characteristics of the crust, and the sensitivity study using MELCOR code was carried out. A stress analysis code using finite element method for the reactor vessel lower head failure analysis was developed and the effect by gap formation between molten corium and vessel was analyzed. Through the international program of PHEBUS-FP and participation in the software development, the study on fission products release and transportation in the software development, the study on fission products release and transportation and aerosol deposition were performed. The system for severe accident analysis codes, CONTAIN and MELCOR codes etc., under the cooperation with USNRC were also established by installing in workstation and applying to experimental results and real plants. (author). 116 refs., 31 tabs., 59 figs

  17. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  18. Hydrogen control technical basis report for severe accident management guidance

    International Nuclear Information System (INIS)

    Hong, Seong Wan; Kim, J. T.; Kim, S. B.

    2005-04-01

    When EPRI wrote the background report on the severe accident management guideline, there were two kinds of hydrogen systems, the hydrogen recombiners and igniters. Today, advanced nuclear power plants is adopting the passive autocatalytic recombiners and/or igniters. The items on the hydrogen control systems adopted in APWR are considered in Candidate High-level Actions. AICC and DDT in Candidate High-level Actions are considered. In chapter 3, the model on the type and performance of PARs is added in phenomena of accident progress. The technical background on the AICC and DDT occurrence evaluation method is described. The influence of the spray system on the detonation is added

  19. 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)

  20. Transient moisture migration in concrete during severe reactor accidents

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Shiina, Y.

    1984-01-01

    In the most severe hypothetical core heatup accidents in High Temperature Gas Cooled Reactors, the heatup of the concrete reactor vessel can result in gas release from degrading concrete which can ultimately lead to containment building failure. This gas release is largely affected by the moisture migration during concrete heatup. Moisture migration in concrete is also of interest in Light Water Reactors (LWR) core meltdown accidents. Therefore, the general problem of moisture migration in a heated concrete slab has been analyzed, including the effect of water evaporating close to the heated surface and recondensing in cooler regions. Results for early phases of core heatup transients are being given. Their implication on the accident progression is being discussed

  1. Temporary jobs and the severity of workplace accidents.

    Science.gov (United States)

    Picchio, Matteo; van Ours, Jan C

    2017-06-01

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

  2. 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.

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

    International Nuclear Information System (INIS)

    Burgherr, Peter; Hirschberg, Stefan

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. Some outstanding issues in severe accidents containment performance

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2004-01-01

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

  6. 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)

  7. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

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

    1996-01-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

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

    International Nuclear Information System (INIS)

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

    2002-03-01

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

  10. Station blackout at nuclear power plants: Radiological implications for nuclear war

    International Nuclear Information System (INIS)

    Shapiro, C.S.

    1986-12-01

    Recent work on station blackout is reviewed its radiological implications for a nuclear war scenario is explored. The major conclusion is that the effects of radiation from many nuclear weapon detonations in a nuclear war would swamp those from possible reactor accidents that result from station blackout

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

    International Nuclear Information System (INIS)

    2002-01-01

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

  12. Overview of severe accident research at the USNRC

    International Nuclear Information System (INIS)

    Basu, S.; Ader, C.E.

    1999-01-01

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

  13. Safety against releases in severe accidents. Final report

    International Nuclear Information System (INIS)

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

    1997-12-01

    The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au)

  14. 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.

  15. Severe Accident Analysis for Containment Filtered Venting System Design

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Suk; Park, Tong Kyu; Lee, Doo Yong; Lee, Byung Chul [FNC Technology Co. Ltd, Yongin (Korea, Republic of); Lee, Sang Won; Kim, Hyeong Taek [KHNP-Central Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Basic idea of containment venting is to relieve the pressure inside of the containment by establishing a flow path to the external environment. After TMI accident, many countries (Sweden, Germany, France) requires containment venting system like FCVS (filtered containment venting system), which can allow for the release of the over-pressure through a scrubber normally containing water and chemicals to reduce the radioactive material releases to the environment. 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. This study examines the thermodynamic behavior due to different vent strategies for a large PWR during severe accidents for the OPR1000 Korea nuclear power plant. The representative accident scenario is identified and the sensitivity analysis with varying conditions (i. e. vent line size and vent initiation pressure) is conducted by using numerical simulation. The effects of venting during the severe accident with containment pressurization are examined. The accident scenarios are selected by using both of the qualitative judgement and the preliminary calculations and the sensitivity analysis on vent line size and vent initiation timing is conducted. As a result, the general trend of containment behavior due to venting can be found. Summarizing the findings, two conflict trends are found: - The maximum discharged flow rate would be higher as the vent line size and vent opening pressure increases. - The decay heat and the aerosol mass delivered to CFVS would be higher as the vent line size and vent opening pressure decreases. Regarding the flow rate, decay heat and aerosol mass are important factor for CFVS design, it would be necessary to find the optimum design specification with economical and regulatory considerations.

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    Purpose Bus safety is a concern not only in developing countries, but also in the U.S. and Europe. In Denmark, disentangling risk factors that are positively or negatively related to bus accident severity and injury occurrence to bus passengers can contribute to promote safety as an essential...... 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...... 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...

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

    International Nuclear Information System (INIS)

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

    2016-01-01

    Influence of thermal-hydrualic phenomena on severe accident unforlding is beyond question. The present paper supports this statement on two key aspects of a severe accident: preservation of containment integrity and transport of fission products once released from fuel. To illustrate them, the attention is focused on suppression pools performance and, particularly, on some recent findings stemming from authors research of Fukushima scenarios. Gas behvaior at the injection point and its later evolution, potential axial and/or azimuthal stratification of the aqueous body or water saturation state, are some of the processes tha more strongly affect the role of pools as a mass and energy sink. They are described and discussed in detail. (Author)

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

    International Nuclear Information System (INIS)

    Vidard, M.

    1998-01-01

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

  20. Hydrogen mitigation by catalytic recombiners and ignition during severe accidents

    International Nuclear Information System (INIS)

    Rohde, J.; Chakraborty, A.K.; Heitsch, M.; Klein-Hebling, W.

    1994-01-01

    A large amount of hydrogen is expected to be released within a large dry containment of a PWR shortly after the onset of a severe accident, leading to core melting. According to local gas concentrations, turbulence and structural configurations within the containment, the released hydrogen can reach the boundary of deflagration or under certain conditions cause local detonations threatening the containment integrity. During the last few years, several concepts of mitigation have been developed to limit the hydrogen concentrations and extensive efforts have been given to investigate the use of catalytic recombiners as well as the use of deliberate ignition within the contemplated framework of a 'Dual-concept'. Although the recent recommendation of the German Reactor Safety Commission (RSK) foresees the sole application of catalytic recombiners to remove hydrogen during severe accident, a review is planned within two years for the partial and directed additional application of early ignitions or post dilution of the atmosphere of the compartments in conjunction with the recombiners installed. This presentation will review the results of large number of experiments performed both in small scale and large scale to qualify the recombiners. It is also the subject of the presentation to address the requirements for proper and secure functioning of the catalyzers under the existing boundary conditions during the severe accidents. These requirements ask for measures, starting from the proper selection of catalysts, multi purposed catalytic devices and their protection against contamination during the standby condition as well as against aerosol deposition and surface poisoning during the propagation of an accident. A short review of the results to large scale experiments with the combined application of catalytic devices and igniters form also a part of this presentation. (author). 8 refs., 2 tabs., 7 figs

  1. Influence of radiation heat transfer during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-09-15

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

  2. Regulatory analyses for severe accident issues: an example

    International Nuclear Information System (INIS)

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

    1984-09-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

    Cao Xuewu; Wang Zhe; Zhang Yingzhen

    2009-01-01

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

  5. Proceedings of the US Nuclear Regulatory Commission nineteenth water reactor safety information meeting. Volume 2, Severe accident research; Severe accident policy implementation; Accident management

    Energy Technology Data Exchange (ETDEWEB)

    Weiss, A.J. [comp.] [Brookhaven National Lab., Upton, NY (United States)

    1992-04-01

    This three-volume report contains 83 papers out of the 108 that were presented at the Nineteenth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, during the week of October 28--30, 1991. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included 14 different papers presented by researchers from Canada, Germany, France, Japan, Sweden, Taiwan, and USSR. This document, Volume 2, presents papers on: Severe accident research; Severe accident and policy implementation; and Accident management. The individual papers have been cataloged separately.

  6. Leak behavior through EPAs under severe accident conditions

    International Nuclear Information System (INIS)

    Keck, J.D.; Thome, F.V.

    1986-01-01

    The leakage behavior of electrical penetration assemblies (EPAs) is being evaluated by varying the penetration type, manufacturer, and hypothetical temperature and pressure accident profile. Nuclear qualified EPAs were procured 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 PWR, a BWR Mark I drywell, and a BWR Mark III wetwell. The EPA manufacturers were matched to the reactor type based on a survey of EPAs used in reactors. The D.G. O'Brien EPA was chosen for the PWR SAC, the Westinghouse was chosen for the BWR Mark III, and the Conax was chosen for the BWR Mark I. The EPAs were radiation and thermal aged to a 40 year service life before the SAC test was conducted. For the D.G. O'Brien EPA SAC test, while there was no detectable leakage during the SAC test, there was a 0.13 scc/sec leak measured during the post-test inspection. For a Westinghouse EPA SAC test, there was no detectable leakage during or after the SAC test. The Conax EPA SAC test is scheduled for completion in July, 1986

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

    Science.gov (United States)

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

    2017-09-01

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

  8. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  9. Level 2 PSA methodology and severe accident management

    International Nuclear Information System (INIS)

    1997-01-01

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

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

    International Nuclear Information System (INIS)

    Yu, Donghan; Ha, Jaejoo

    1998-01-01

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

  11. Cofrentes NPP activities on PSA and severe accident analysis

    International Nuclear Information System (INIS)

    Suarez, J.; Borondo, L.; Garcia, P.J.

    1996-01-01

    Cofrentes NPP (CNPP) has developed a Level 1 PSA with the following scope: analysis of internal events, with the reactor initially operating at power, internal and external flooding risk analysis; internal fire risk analysis; reliability analysis of the containment heat removal and containment isolation systems. Level 1 CNPP-PSA results reveal that total core damage frequency in CNPP is less than other similar BWR/6 plants. The CNPP-PSA related activities and applications being carried out currently are: adjusting of MAAP 3.0B, revision 10, on VAX and PC; acquisition of MAAP 4; development of Level1/Level2-PSA interface; seismic site categorization for the IPEEE; prioritization of motor operated valves related to GL-89/10, complementary analysis for exemption to some 10CFR50 App. J requirements; Q-List grading; reliability-centered maintenance; maintenance rule support; on-line maintenance support, off-line risk-monitor development, PSA applicability to the 10CFR50 App. R requirements, analysis of the frequency of mis-oriented fuel bundle event, etc. About severe accident management, CNPP, as part of the Spanish-BWROG, is currently analyzing the generic products of the US-BWROG AMG in order to generate their specific ones. Also, in this group BWR, the development of tools to simulate accident scenarios beyond core damage will be studied and a training process oriented to warrant the optimum use of new EOP/AMG in accident scenarios will be implemented

  12. Station blackout consequences and responses for Daya Bay NPP

    International Nuclear Information System (INIS)

    Shen Rugang

    1995-01-01

    The paper describes briefly the phenomena, consequences of station blackout and responses taken in Daya Bay Nuclear Power Plant which include technical counter measures and accident procedures. Finally, a brief analysis and calculation of residual risk due to complete loss of electric power supplies are also introduced

  13. A physical tool for severe accident mitigation studies

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-01

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

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

    International Nuclear Information System (INIS)

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

    1992-02-01

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

  15. Fuel behaviour in the case of severe accidents and potential ATF designs. Fuel Behavior in Severe Accidents and Potential Accident Tolerance Fuel Designs

    International Nuclear Information System (INIS)

    Cheng, Bo

    2013-01-01

    This presentation reviews the conditions of fuel rods under severe loss of coolant conditions, approaches that may increase coping time for plant operators to recover, requirements of advanced fuel cladding to increase tolerance in accident conditions, potential candidate alloys for accident-tolerant fuel cladding and a novel design of molybdenum (Mo) -based fuel cladding. The current Zr-alloy fuel cladding will lose all its mechanical strength at 750-800 deg. C, and will react rapidly with high-pressure steam, producing significant hydrogen and exothermic heat at 700-1000 deg. C. The metallurgical properties of Zr make it unlikely that modifications of the Zr-alloy will improve the behaviour of Zr-alloys at temperatures relevant to severe accidents. The Mo-based fuel cladding is designed to (1) maintain fuel rod integrity, and reduce the release rate of hydrogen and exothermic heat in accident conditions at 1200-1500 deg. C. The EPRI research has thus far completed the design concepts, demonstration of feasibility of producing very thin wall (0.2 mm) Mo tubes. The feasibility of depositing a protective coating using various techniques has also been demonstrated. Demonstration of forming composite Mo-based cladding via mechanical reduction has been planned

  16. Assessment of ICARE/CATHARE V1 Severe Accident Code

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  17. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. Importance of prototypic-corium experiments for severe accident research

    International Nuclear Information System (INIS)

    Piluso, P.; Journeau, C.; Cognet, G.; Magallon, D.; Seiler, J.M.

    2001-01-01

    In case of a severe accident in a nuclear reactor, very complex physical and chemical phenomena would occur. Parallel to the development of mechanistic and scenario codes, experiments are needed to determine key phenomena and coupling, develop and qualify specific models, validate codes. Experiments with prototypic corium are performed to check the results obtained with corium-simulant materials and identify possible differences. In this context CEA has undertaken a large program on severe accidents with prototypic corium. In this paper, we discuss some specificities of the prototypic corium: 1) Spreading: experiments with simulant mixtures and prototypic corium performed in the VULCANO facility showed a behaviour involving gas formation during melt spreading. 2) Corium pool: the presence of miscibility gap in the U-Zr-O ternary system for liquid phases and the high density of uranium oxides affect solidification paths, stratification and/or macro-segregation. 3) Corium concrete interaction: the possible reactions between uranium oxide and concrete oxides are specific in terms of thermodynamics and kinetics. For instance, the limited solubility of uranium in zircon can lead to the formation of the solid solution called ''chernobylite'' (U x ,Zr 1-x )SiO 4 which is important for the long term behaviour (fission product release, handling,..) of solidified corium. 4) Fuel Coolant Interaction: experiments in the KROTOS facility have shown important differences of behaviour between molten alumina and molten 80% wt UO 2 + 20% wt ZrO 2 , the latter inducing less violent explosions than the former

  19. Application of severe accident code with graphic interface

    International Nuclear Information System (INIS)

    Villegas Gonzalez, O.A.

    1993-01-01

    The RELAP5/SCDAP/MOD3 code version 7af, has been installed on the CONVEX/C-220 computer under Unix operative system in the Severe Accident Unit (UAG), of the Consejo de Seguridad Nuclear (CSN). The objectives are to carry out calculations involving thermalhydraulics of the Reactor Coolant System (RCS) and core damage evolution and fission products release and transport during severe accident in plants and experimental facilities like PHEBUS/FP. The Nuclear Plant Analyzer display software (NPA) version 1.2, has been also installed on. Improvements have been performanced into NPA software involving implementation to CONVEX environment. Algorithms involving some dynamics variables has been also improved. Finally, NPA has been adapted in part, to the powerful graphics capacity CONVEX/CX MOTIF, which corresponds to an implementation of OSF/MOTIF product for CONVEX supercomputers build on CX-Windows. CX/MOTIF environment involves various components as: The toolkit, window manager and users interfaces that permits a quick access to lower level of X-Window. The code was linked to NPA software in order to adquire a training and knowledge of the graphic software package and to obtain an interactive animated response display of the different applications that UAG has foreseed to carry out with RELAP5/SCDAP/MOD3 as same as operator-like actions as means for controlling the simulation system. (orig.)

  20. Advances in operational safety and severe accident research

    International Nuclear Information System (INIS)

    Simola, K.

    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)

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. Large blackouts in North America: Historical trends and policy implications

    Energy Technology Data Exchange (ETDEWEB)

    Hines, Paul [School of Engineering, 301 Votey Hall, University of Vermont, 33 Colchester Avenue, Burlington, VT 05405 (United States); Apt, Jay [Carnegie Mellon Electricity Industry Center, Department of Engineering and Public Policy and Tepper School of Business, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States); Talukdar, Sarosh [Carnegie Mellon Electricity Industry Center, Department of Engineering and Public Policy and Department of Electrical and Computer Engineering, Carnegie Mellon University, 5000 Forbes Avenue, Pittsburgh, PA 15213 (United States)

    2009-12-15

    Using data from the North American Electric Reliability Council (NERC) for 1984-2006, we find several trends. We find that the frequency of large blackouts in the United States has not decreased over time, that there is a statistically significant increase in blackout frequency during peak hours of the day and during late summer and mid-winter months (although non-storm-related risk is nearly constant through the year) and that there is strong statistical support for the previously observed power-law statistical relationship between blackout size and frequency. We do not find that blackout sizes and blackout durations are significantly correlated. These trends hold even after controlling for increasing demand and population and after eliminating small events, for which the data may be skewed by spotty reporting. Trends in blackout occurrences, such as those observed in the North American data, have important implications for those who make investment and policy decisions in the electricity industry. We provide a number of examples that illustrate how these trends can inform benefit-cost analysis calculations. Also, following procedures used in natural disaster planning we use the observed statistical trends to calculate the size of the 100-year blackout, which for North America is 186,000 MW. (author)

  3. Large blackouts in North America: Historical trends and policy implications

    International Nuclear Information System (INIS)

    Hines, Paul; Apt, Jay; Talukdar, Sarosh

    2009-01-01

    Using data from the North American Electric Reliability Council (NERC) for 1984-2006, we find several trends. We find that the frequency of large blackouts in the United States has not decreased over time, that there is a statistically significant increase in blackout frequency during peak hours of the day and during late summer and mid-winter months (although non-storm-related risk is nearly constant through the year) and that there is strong statistical support for the previously observed power-law statistical relationship between blackout size and frequency. We do not find that blackout sizes and blackout durations are significantly correlated. These trends hold even after controlling for increasing demand and population and after eliminating small events, for which the data may be skewed by spotty reporting. Trends in blackout occurrences, such as those observed in the North American data, have important implications for those who make investment and policy decisions in the electricity industry. We provide a number of examples that illustrate how these trends can inform benefit-cost analysis calculations. Also, following procedures used in natural disaster planning we use the observed statistical trends to calculate the size of the 100-year blackout, which for North America is 186,000 MW.

  4. Human reliability analysis for venting a BWR Mark I during a severe accident

    International Nuclear Information System (INIS)

    Nelson, W.R.; Blackman, H.S.

    1986-01-01

    A Human Reliability Analysis (HRA) was performed for the operator actions necessary to achieve containment venting for the Peach Bottom Atomic Power Station. This study was funded by the United States Nuclear Regulatory Commission (USNRC) and performed by the Idaho National Engineering Laboratory (INEL). The goal of the analysis was to estimate Human Error Probabilities (HEPs) to determine the likelihood that operators would fail to complete the venting process. The analysis was performed for two generic accident sequences: anticipated transient without scram (ATWS) and station blackout. Two major methods were used to estimate the HEPs: Technique for Human Error rate Prediction (THERP) and Success Likelihood Index Methodology (SLIM). For the ATWS scenarios analyzed, the calculated HEPs ranged from 0.23 to 0.35, depending on the number of vent paths that are required to reduce the containment pressure. It should be noted that the confidence bounds around these HEPs are large, However, even when considering the large confidence range, the failure probabilities are larger than what is typical for normal operator actions. For station blackout, the HEP is 1.0, resulting from the dangerous environmental conditions that are present, assuming that plant management would not deliberately expose personnel to a potentially fatal environment. These results are based on the analysis of draft procedures for containment venting. It is probable that careful revision of the procedures could reduce the human error probabilities

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-04-15

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

  6. Shipping container response to three severe railway accident scenarios

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  7. Shipping container response to three severe railway accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-04-01

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

  8. Specific features of RBMK severe accidents progression and approach to the accident management

    International Nuclear Information System (INIS)

    Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

    2001-01-01

    Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

  9. BNL severe accident sequence experiments and analysis program

    International Nuclear Information System (INIS)

    Greene, G.A.; Ginsberg, T.; Tutu, N.K.

    1985-01-01

    Analyses of LWR degraded core accidents require mathematical characterization of two major sources of pressure and temperature loading on the reactor containment buildings: (1) steam generation from core debris-water thermal interactions and (2) molten core-concrete interactions. Experiments are in progress at BNL in support of analytical model development related to aspects of the above containment loading mechanisms. The work supports development and evaluation of the CORCON, MARCH, CONTAIN and MEDICI computer under development at other NRC-contractor laboratories. The thermal-hydraulic behavior of hot debris located within the reactor core region upon sudden introduction of cooling water is being investigated in a joint experimental and analytical program. This work supports development and evaluation of the SCDAP computer code being developed at EG and G to characterize in-vessel severe core damage accident sequences. Progress is described in the two areas of: 1) core debris thermal-hydraulic phenomenology and 2) heat transfer in core-concrete interactions

  10. Probabilistic Assessment of Severe Accident Consequence in West Bangka

    Science.gov (United States)

    Sunarko; Su'ud, Zaki

    2017-07-01

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

  11. A hypothetical severe reactor accident in Sosnovyj Bor, Russia

    International Nuclear Information System (INIS)

    Lahtinen, J.; Toivonen, H.; Poellaenen, R.; Nordlund, G.

    1993-12-01

    Individual doses and short-term radiological consequences from a hypothetical severe accident at the Russian nuclear power plant in Sosnovyj Bor were estimated for two sites in Finland. The sites are Kotka, located 140 km from the plant, and Helsinki, 220 km from the plant. The release was assumed to start immediately after the shutdown of the reactor (a 1000 MW RBMK unit) which had been operating at nominal power level for a long time. An effective release height of 500 m was assumed. The prevailing meteorological conditions during the release were taken to present the situation typical of the area (effective wind speed 9 m/s, neutral dispersion conditions). The release fractions applied in the study were of the same order as in the Chernobyl accident, i.e. 100% for noble gases, 60% for iodines, 40% for cesium and 1-10% for other radiologically important nuclides. The release was assumed to last 24 hours. However, half of the nuclides were released during the first hour. No attention was paid to the actual sequence of events that could lead to such release characteristics and time behaviour. The concentration and dose calculations were performed with a modified version of the computer code OIVA developed in Finnish Centre for Radiation and Nuclear Safety. Inhalation dose and external doses from the release plume and from the deposited activity were calculated for adults only, and no sheltering was considered. (11 refs., 4 figs., 6 tabs.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1986-09-01

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

  13. 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.

  14. 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.

  15. 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.

  16. Aerosol behavior in the reactor containment building during severe accident

    International Nuclear Information System (INIS)

    Berthion, Y.; Lhiaubet, G.; Gauvain, J.

    1984-07-01

    Thermohydraulic behavior inside a PWR containment during severe accident depends on decay heat transferred to the sump water by aerosol gravitational settling and deposition. Conversely, aerosol behavior depends on thermal hydraulic conditions, especially atmosphere moisture for soluble aerosol GsI, and CsOH. Therefore, a small iterative procedure between thermo-hydraulic and aerosol calculations has been performed in order to evaluate the importance of this coupling between the two phenomena. In this paper, it is shown that with this procedure and using our codes JERICHO, RICOCHET and AEROSOLS/B1, the steam condensation on aerosols is an important phenomenon for a correct estimation of the attenuation factor of the suspended mass of aerosols in the airborne of the containment. Then, we have a more realistic assessment of the source term released by the containment

  17. 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.

  18. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

    Goetz, W.W.J.; Seebregts, A.J.; Bedford, T.J.

    1996-08-01

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

  19. Seismic isolation of plants at risk of a severe accident

    International Nuclear Information System (INIS)

    Forni, Massimo

    2015-01-01

    More and more devastating earthquakes struck every year our planet. Many of these, though occurring in areas considered at high risk of earthquakes, far exceed the levels required by law. The industrial plants subjected to risk of severe accident, in particular petrochemical and nuclear power plants, are particularly exposed to this risk because of the number and the complexity of the structures and critical components of which they are composed. For this type of structures, anti-seismic techniques able to provide complete protection, even in case of unforeseen events, are needed. Seismic isolation is certainly the most promising technology of modern antiseismic as it allows not only to significantly reduce the dynamic load acting on the structures in case of seismic attack, but to provide safety margins against violent earthquakes, exceeding the assumed maximum design limit. [it

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  1. Importance of prototypic-corium experiments for severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Piluso, P.; Journeau, C.; Cognet, G. [CEA Cadarache, 13 - Saint Paul lez Durance (France); Magallon, D. [European Commission, JRC Institute for Advanced Material, Petten (Netherlands); Seiler, J.M. [CEA Grenoble, 38 (France)

    2001-07-01

    In case of a severe accident in a nuclear reactor, very complex physical and chemical phenomena would occur. Parallel to the development of mechanistic and scenario codes, experiments are needed to determine key phenomena and coupling, develop and qualify specific models, validate codes. Experiments with prototypic corium are performed to check the results obtained with corium-simulant materials and identify possible differences. In this context CEA has undertaken a large program on severe accidents with prototypic corium. In this paper, we discuss some specificities of the prototypic corium: 1) Spreading: experiments with simulant mixtures and prototypic corium performed in the VULCANO facility showed a behaviour involving gas formation during melt spreading. 2) Corium pool: the presence of miscibility gap in the U-Zr-O ternary system for liquid phases and the high density of uranium oxides affect solidification paths, stratification and/or macro-segregation. 3) Corium concrete interaction: the possible reactions between uranium oxide and concrete oxides are specific in terms of thermodynamics and kinetics. For instance, the limited solubility of uranium in zircon can lead to the formation of the solid solution called ''chernobylite'' (U{sub x},Zr{sub 1-x})SiO{sub 4} which is important for the long term behaviour (fission product release, handling,..) of solidified corium. 4) Fuel Coolant Interaction: experiments in the KROTOS facility have shown important differences of behaviour between molten alumina and molten 80%{sub wt} UO{sub 2} + 20%{sub wt} ZrO{sub 2}, the latter inducing less violent explosions than the former.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  3. A methodology for the transfer of probabilities between accident severity categories

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Neuhauser, K.S.

    1992-01-01

    Evaluation of the radiological risks of accidents involving vehicles transporting radioactive materials requires consideration of both accident probability and consequences. The probability that an accident will occur may be estimated from historical accident data for the given mode of transport. In addition to an overall accident rate, information regarding accident severity and the resulting package environments across the range of all credible accidents is needed to determine the potential for a release of radioactive material from the package or for an increase in direct radiation from the package caused by damage to packaging shielding. This information is usually obtained from a variety of sources such as historical data, experimental data, analyses of accident and package environments, and expert opinion. The consequences of an accident depend on a number of factors including the type, quantity, and physical form of radioactive material being transported; the response of the package to accident environments; the fraction of material released from the package; and the dispersion of any released material. One approach for the classification and treatment of transportation accidents in risk analysis divides the complete range of critical accident environments resulting from all credible accidents into some number of accident-severity categories. The types of accident environments that a package may be subjected to in transportation are often classified into the following five groups: impact, fire, crush, puncture, and immersion. A open-quotes criticalclose quotes accident environment is one of a type that could present a plausible threat to a package. Each severity category represents a portion of all credible accidents, and the total of all severity categories covers the complete range of critical accident environments. This approach is used in the risk assessment codes RADTRAN (Neuhauser and Kanipe 1992) and INTERTRAN (Ericsson and Elert 1983)

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

    International Nuclear Information System (INIS)

    Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

    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

  5. 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.

  6. Stepwise Reconstruction PROGRAMME of Bohunice V 1 NPP with Regard to SEVERE ACCIDENT Prevention and Mitigation

    International Nuclear Information System (INIS)

    Kuehne, M.; Kerak, M.; Severa, M.

    1997-01-01

    Accident Management means - generally speaking - all measures taken to 'rectum the plant to a safe state and mitigate the consequences of accidents' in the design basis as well as the beyond design basis realm. Sometimes accident management is interpreted as reforming in particular to those actions which are taken to cope with beyond design basis accidents, i.e. in very unlikely situations. The objectives of the measures depend on the category of the accident: design basis accident: * to keep systems and the plant within the licensed limits beyond design basis accident: * to prevent severe core damage, especially at high pressure in the reactor pressurize vessel and or * to mitigate the consequences of severe accidents for the environment The system use is different: design basis accident: * operation of systems within design limits beyond design basis accident: * best use of all available systems even beyond their design limits assuming best estimate conditions for components and systems. With respect to the different system uses it is expedient to distinguish between measures to control accidents under design basis accident conditions or to manage beyond design basis accidents. The evaluation of a safety concept should give precedence to the systems fundamental roles in the design basis realm. beyond design basis -accident management should be seen as an additional, last-resort tool to cope with scenarios in the beyond design basis realm and should therefore be treated separately as a supplement to the 'normal' design basis measures

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

    Directory of Open Access Journals (Sweden)

    Jantanee Dumrak

    2013-12-01

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

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

    International Nuclear Information System (INIS)

    Park, Jae Hong; Jeun Gyoo Dong; Bang, Kwang Hyun; Park, Seh In; Lim, Jae Hyuck; Park, Seong Yong; Back, Hyung Hmm

    1998-03-01

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

  9. 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)

  10. 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.

  11. SARNET: An European cooperative effort on LWR severe accident research

    International Nuclear Information System (INIS)

    Micaelli, Jean-Claude; Van Dorsselaere, Jean-Pierre; Chaumont, Bernard; Adroguer, Bernard; Haste, Tim; Bonnet, Jean-Michel; Meyer, Leonhard; Beraha, David; Trambauer, Klaus; Annunziato, Alessandro; Sehgal, Raj

    2006-01-01

    49 organisations network in SARNET (Severe Accident Research and management NETwork) their capacities of research in order to resolve the most important remaining uncertainties and safety issues for enhancing, in regard of Severe Accidents (SA), the safety of existing and future Nuclear Power Plants (NPPs). This project has been defined bearing in mind the necessity to optimise the use of the available means and to constitute sustainable research groups. SARNET tackles the fragmentation that exists between the different R and D national programmes, notably in defining common research programmes and developing common computer tools and methodologies for safety assessment. SARNET comprises most of the actors involved in SA research in Europe. To reach these objectives, all the organizations networked in SARNET contribute to a so-called Joint Programme of Activities (JPA), which can be broken in 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 code (integral computer code used to predict the NPP behaviour during a postulated SA), which capitalizes in terms of physical models the knowledge produced within SARNET; - Developing Scientific Databases, in which all the results of research programmes are stored; - 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. A few organizations are covering a wide range of competences though not complete, whereas others are specialized in very specific areas and thus complementarities are developing. The critical mass of competence for performing experiments needed in the

  12. Relationship between accident severity and full-scale crash test. Volume II, Appendices

    Science.gov (United States)

    1984-08-01

    Available accident files are used to generate a 4l2-accident data base of guardrail impacts. This base is analyzed to develop a statistical model for predicting accident severity index (ASI) as a function of vehicle type or weight, impact speed, and ...

  13. Relationship between accident severity and full-scale crash test. Volume I, Technical research effort

    Science.gov (United States)

    1984-08-01

    Available accident files are used to generate a 4l2-accident data base of guardrail impacts. This base is analyzed to develop a statistical model for predicting accident severity index (ASI) as a function of vehicle type or weight, impact speed, and ...

  14. Chemical factors affecting fission product transport in severe LMFBR accidents

    International Nuclear Information System (INIS)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly

  15. Chemical factors affecting fission product transport in severe LMFBR accidents

    Energy Technology Data Exchange (ETDEWEB)

    Wichner, R.P.; Jolley, R.L.; Gat, U.; Rodgers, B.R.

    1984-10-01

    This study was performed as a part of a larger evaluation effort on LMFBR accident, source-term estimation. Purpose was to provide basic chemical information regarding fission product, sodium coolant, and structural material interactions required to perform estimation of fission product transport under LMFBR accident conditions. Emphasis was placed on conditions within the reactor vessel; containment vessel conditions are discussed only briefly.

  16. A methodology for the transfer of probabilities between accident severity categories

    International Nuclear Information System (INIS)

    Whitlow, J.D.; Neuhauser, K.S.

    1993-01-01

    This paper will describe a methodology which has been developed to allow accident probabilities associated with one severity category scheme to be transferred to another severity category scheme, permitting some comparisons of different studies at the category level. In this methodology, the severity category schemes to be compared are mapped onto a common set of axes. The axes represent critical accident environments (e.g., impact, thermal, crush, puncture) and indicate the range of accident parameters from zero (no accident) to the most sever credible forces. The choice of critical accident environments for the axes depends on the package being transported and the mode of transportation. The accident probabilities associated with one scheme are then transferred to the other scheme. This transfer of category probabilities is based on the relationships of the critical accident parameters to probability of occurrence. The methodology can be employed to transfer any quantity between category schemes if the appropriate supporting information is available. (J.P.N.)

  17. Relap5 Analysis of Processes in Reactor Cooling Circuit and Reactor Cavity in Case of Station Blackout in RBMK-1500

    International Nuclear Information System (INIS)

    Kaliatka, A.

    2007-01-01

    Ignalina NPP is equipped with channel-type boiling-water graphite-moderated reactor RBMK-1500. Results of the level-1 probabilistic safety assessment of the Ignalina NPP have shown that in topography of the risk, the transients with failure of long-term core cooling other than LOCA are the main contributors to the core damage frequency. The total loss of off-site power with a failure to start any diesel generator, that is station blackout, is the event which could lead to the loss of long-term core cooling. Such accident could lead to multiple ruptures of fuel channels with severe consequences and should be analyzed in order to estimate the timing of the key events and the possibilities for accident management. This paper presents the results of the analysis of station blackout at Ignalina NPP. Analysis was performed using thermal-hydraulic state-of-the-art RELAP5/MOD3.2 code. The response of reactor cooling system and the processes in the reactor cavity and its venting system in case of a few fuel-channel ruptures due to overheating were demonstrated. The possible measures for prevention of the development of this beyond design basis accident (BDBA) to a severe accident are discussed

  18. Addressing Hydrogen Risks During Severe Accidents - EDF Approach

    International Nuclear Information System (INIS)

    Guieu, Serge

    2001-01-01

    Hydrogen mitigation systems for French PWRs have not been designed to cope with hydrogen release rates and quantities equivalent to those contemplated if some Severe Accident sequences were to occur. Prior to making any decision on the need for addressing this issue, EDF has looked at available alternatives, decided which one could be a reasonable candidate if further studies showed that the problem had to be addressed, and started R and D programmes aiming at the validation of the preferred candidate. This paper describes the process followed to evaluate the need for enhancing the capability of hydrogen mitigation measures, indicates which candidates were initially considered for mitigating the hydrogen risk and gives the rationales for selecting Passive Auto-catalytic Recombiners (PARS) as the preferred alternative. The PAR validation process is described and the main results confirming their capability are commented. Finally, insights on future on-site implementation are given. As a general conclusion, all the EDF 'efficiency tests programmes' described, in addition with results of other similar tests programmes performed in other countries (in particular: Germany and Canada) do not challenge the capability of PARS to operate in severe accidents representative conditions. Concerning the risk for a PAR to initiate a deflagration, EDF tests have shown that this risk is depending both on hydrogen and steam contents. The risk is more accurate when steam content is low. Roughly speaking, the hydrogen content limit, in presence of steam, to avoid this phenomenon is in the range of 8 vol %. In addition it has to be noted that this 'deflagration problem', does not challenge the interest to install PARS because: - in the containment of present NPPs, numerous ignition sources are still present, - in case of initiation of a deflagration by a PAR, it is likely that this deflagration will appear with an hydrogen amount in the containment much lower than without PARs

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Frid, W.

    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

  1. Design Provisions for Station Blackout at Nuclear Power Plants

    International Nuclear Information System (INIS)

    Duchac, Alexander

    2015-01-01

    A station blackout (SBO) is generally known as 'a plant condition with complete loss of all alternating current (AC) power from off-site sources, from the main generator and from standby AC power sources important to safety to the essential and nonessential switchgear buses. Direct current (DC) power supplies and un-interruptible AC power supplies may be available as long as batteries can supply the loads. Alternate AC power supplies are available'. A draft Safety Guide DS 430 'Design of Electrical Power Systems for Nuclear Power Plants' provides recommendations regarding the implementation of Specific Safety Requirements: Design: Requirement 68 for emergency power systems. The Safety Guide outlines several design measures which are possible as a means of increasing the capability of the electrical power systems to cope with a station blackout, without providing detailed implementation guidance. A committee of international experts and advisors from numerous countries is currently working on an IAEA Technical Document (TECDOC) whose objective is to provide a common international technical basis from which the various criteria for SBO events need to be established, to support operation under design basis and design extension conditions (DEC) at nuclear power plants, to document in a comprehensive manner, all relevant aspects of SBO events at NPPs, and to outline critical issues which reflect the lessons learned from the Fukushima Dai-ichi accident. This paper discusses the commonly encountered difficulties associated with establishing the SBO criteria, shares the best practices, and current strategies used in the design and implementation of SBO provisions and outline the structure of the IAEA's SBO TECDOC under development. (author)

  2. SWR-1000 concept on control of severe accidents

    International Nuclear Information System (INIS)

    Meyer, P.J.

    1998-01-01

    It is essential for the SWR-1000 probabilistic safety concept to consider the results from experiments and reliability system failure within the probabilistic safety analyses for passive systems. Active and passive safety features together reduce the probability of the occurrence of beyond design basis accidents in order to limit their consequences in accordance with the German law. As a reference case we analyzed the most probable core melt accident sequence with a very conservative assumption. An initial event, stuck open of safety and relief valves without the probability of active and passive feeding systems of the pressure vessel, was considered. Other sequences of the loss of coolant accidents lead to lower probability

  3. Neutronic analysis of LMFBRs during severe core disruptive accidents

    International Nuclear Information System (INIS)

    Tomlinson, E.T.

    1979-01-01

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

  4. Prevention of heavy missiles during severe PWR accidents

    International Nuclear Information System (INIS)

    Krieg, R.

    1994-01-01

    For future pressurized water reactors, which should be designed against core melt down accidents, missiles generated inside the containment present a severe problem for its integrity. The masses and geometries of the missiles as well as their velocities may vary to a great extend. Therefore, a reliable proof of the containment integrity is very difficult. To overcome this problem the potential sources of missiles are discussed. In section 5 it is concluded that the generation of heavy missiles must be prevented. Steam explosions must not damage the reactor vessel head. Thus fragments of the head cannot become missiles endangering the containment shell. Furthermore, during a melt-through failure of the reactor vessel under high pressure the resulting forces must not catapult the whole vessel against the containment shell. Only missiles caused by hydrogen explosions might be tolerable, but shielding structures which protect the containment shell might be required. Here further investigations are necessary. Finally, measures are described showing that the generation of heavy missiles can indeed be prevented. In section 6 investigations are explained which will confirm the strength of the reactor vessel head. In section 7 a device is discussed keeping the fragments of a failing reactor vessel at its place. (author). 12 refs., 8 figs

  5. 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.).

  6. 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.

  7. 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.

  8. Severe accident progression perspectives based on IPE results

    International Nuclear Information System (INIS)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T.; Drouin, M.

    1996-01-01

    Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

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

    International Nuclear Information System (INIS)

    1990-12-01

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-03-15

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

  12. Human factors review for nuclear power plant severe accident sequence analysis

    International Nuclear Information System (INIS)

    Krois, P.A.; Haas, P.M.

    1985-01-01

    The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release

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

    International Nuclear Information System (INIS)

    Camous, F.; Jacq, F.; Chatelard, P.

    1997-01-01

    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

  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. 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.

  16. 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.)

  17. Severe accident analysis using MARCH 1.0 code

    International Nuclear Information System (INIS)

    Guimaraes, A.C.F.

    1987-09-01

    The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

  18. 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.

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    at three hospitals located at Funen in the period between 2002 and 2009. Furthermore, two large-scale questionnaires were developed and administered. The first questionnaire was administered among bicyclist in Denmark and investigates the behavioural reasons for reporting traffic accidents. The second...... of the reasons for heterogeneity has been made, which in the end may lead to devising policy measures (Paper 1). 3) A connection between the occurrence probability of trauma type and crash, vehicle and person characteristics exists (Paper 2). 4) The attitudes that accident reporting is useless are found...... 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...

  20. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM

    International Nuclear Information System (INIS)

    Castillo G, F.

    2015-01-01

    In this thesis work, it was developed a model of the vacuum breaker valves and down comers for a BWR Mark II primary containment for the code RELAP/SCDAPSIM Mod. 3.4. This code was used to simulate a Station Blackout (Sbo) that evolves to a severe accident scenario. To accomplish this task, the vacuum breaker valves and down comers were included in a simplified model of the primary containment that includes both wet well and dry well, which was coupled with a model of the Nuclear Steam Supply System (NSSS), in order to study the behavior of the primary containment during the evolution of the accident scenario. In the analysis of the results of the simulation, the behavior of the wet well and dry well during the event was particularly monitored, by analyzing the evolution of temperature and pressure profiles in such volumes, this to determine the impact of the inclusion of the breaker vacuum valves and down comers. The results show that the effect of this extension of the model is that more conservative results are obtained, i.e., higher pressures are reached in both wet well and dry well than when it is used a containment model that does not include neither the vacuum valves nor the down comers. The most relevant results obtained show that the Rcic alone is able to keep the core fully covered, but even in such a case, it evaporates about 15% of the initial inventory of liquid water in the Pressure Suppression Pool (Psp). When the Rcic operation is lost, 20% more of the liquid water inventory in the Psp is further reduced within four to twelve hours (approximately), time at which the simulation crashed. Besides, there is a significant increase of pressure in the containment. As the accident evolves, the pressure in the containment continues increasing, but there is still considerable margin to reach the design pressure of the containment. At the end of the simulation, the results show a gauge pressure value of 224,550 Pa in the Psp and 187,482 Pa in the wet well

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

    International Nuclear Information System (INIS)

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

    2000-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

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

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Jae, Moosung

    2015-01-01

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

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

    International Nuclear Information System (INIS)

    L'Homme, A.; Pelce, J.

    1986-07-01

    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 [fr

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

    International Nuclear Information System (INIS)

    2011-07-01

    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.

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

    International Nuclear Information System (INIS)

    1994-01-01

    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. MASCA, In-vessel phenomena during severe accidents

    International Nuclear Information System (INIS)

    2007-01-01

    Description: The MASCA Project was a follow-up of the RASPLAV Project and investigated in-vessel phenomena during a severe accident. In particular, it addressed the influence of the chemical composition of the molten corium on the heat transfer to the pressure vessel environment. The project addressed this by investigating stratification phenomena of the molten pool and the partitioning of fission products (FP) within the different layers of the melt. The project was scheduled to be completed in July 2003, but it was continued until 2006 under the MACS-2 Project, given the experimental needs that still exist and the quality of the experimental work done so far. The tests aimed to resolve remaining uncertainties about the heat load on the reactor vessel and thus the possibility of retaining the melt in the vessel. These uncertainties are mainly associated with scaling effects and coupling between the thermal-hydraulic and chemical behaviour of the melt. Supporting experiments and analyses - in addition to helping understand key in-vessel phenomena - facilitated a consistent interpretation of the results. The experiments were carried out with corium compositions prototypical of power reactors which use iron and steel materials. The MASCA experimental goal was achieved through corium tests of different scale, and was complemented by pre- and post-test analyses and development of computational models. Additional measurements of thermo-physical properties of the melts such as density, thermal conductivity and liquidus-solidus temperatures considerably expanded the material properties data obtained during the RASPLAV Project. The major goals of the MASCA Project were to: - Investigate the influence of chemical behaviour on heat transfer in stratified molten pools of prototypical compositions; - Investigate FP behaviour in a molten pool and in particular: Partitioning of FP between layers in case of stratification; Partitioning of FP between phases during melting and

  8. 78 FR 21275 - Station Blackout Mitigation Strategies

    Science.gov (United States)

    2013-04-10

    ... 52 [NRC-2011-0299] RIN 3150-AJ08 Station Blackout Mitigation Strategies AGENCY: Nuclear Regulatory... Regulations (10 CFR) to incorporate requirements involving station blackout mitigation strategies, the NRC is... regulatory basis to incorporate requirements involving station blackout mitigation strategies (ADAMS...

  9. 78 FR 44035 - Station Blackout Mitigation Strategies

    Science.gov (United States)

    2013-07-23

    ... applicants' station blackout mitigation strategies. The issuance of this regulatory basis document is one [email protected] . The regulatory basis document, ``Station Blackout Mitigation Strategies,'' is available in... incorporate requirements involving station blackout mitigation strategies (SBOMS), the NRC is making documents...

  10. Assessment of impact of a severe accident at nuclear power plant of Angra dos Reis with release of radionuclides to the atmosphere

    International Nuclear Information System (INIS)

    Aguiar, Andre Silva de

    2015-01-01

    This study had as purpose the assess the impact of a severe accident, and also analyze the dispersion of 131 I in the atmosphere, so that, through concentrating and inhaling dose of the plume, were possible to verify if the results are in accordance with the indicated data by the Plan of Emergency of the CNAAA regarding the Impact Zone and Control. This exercise was performed with the aid of an atmospheric model and a dispersion where to atmospheric modeling we used the data coupling WRF / CALMET and of dispersion, CALPUFF. The suggested accident consists of a Station Blackout at Nuclear Power of Angra (Unit 1), where through the total core involvement, will release 100% of the 131 I to the atmosphere. The value of the total activity in the nucleus to this radionuclide is 7.44 x 1017 Bq, that is relative on the sixth day of burning. This activity will be released through the chimney at a rate in Bq/s in the scenario of 12, 24, 48 and 72 hours of release. Applying the model in the proposed scenario, it is verified that the plume has concentrations of the order of 1020 Bq/m³ and dose of about 108 Sv whose value is beyond of the presented by Eletronuclear in your current emergency plan. (author)

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

    Science.gov (United States)

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

    2015-06-01

    Radiation protection issues on preparedness and response for a severe nuclear accident are discussed in this paper based on the experiences following the accident at Fukushima Daiichi nuclear power plant. The criteria for use in nuclear emergencies in the Japanese emergency preparedness guide were based on the recommendations of International Commission of Radiological Protection (ICRP) Publications 60 and 63. Although the decision-making process for implementing protective actions relied heavily on computer-based predictive models prior to the accident, urgent protective actions, such as evacuation and sheltering, were implemented effectively based on the plant conditions. As there were no recommendations and criteria for long-term protective actions in the emergency preparedness guide, the recommendations of ICRP Publications 103, 109, and 111 were taken into consideration in determining the temporary relocation of inhabitants of heavily contaminated areas. These recommendations were very useful in deciding the emergency protective actions to take in the early stages of the Fukushima accident. However, some suggestions have been made for improving emergency preparedness and response in the early stages of a severe nuclear accident. © The Chartered Institution of Building Services Engineers 2014.

  12. Guidelines for calculation of atmospheric dispersion and radiological consequences of design basis reactor accidents - Severe accident calculation guidelines, EPR

    International Nuclear Information System (INIS)

    Martens, R.; Schmitz, B.M.; Horn, M.

    1999-01-01

    The activities carried out within the (reduced) project period (1. Sept. until 31. Dec. 1998) for coordinated harmonization between France and Germany, of guidelines for calculation of the radiological consequences of a severe reactor accident, are summarized. (orig./CB) [de

  13. Computer chaos and the blackout

    CERN Multimedia

    Malik, Rex

    1971-01-01

    A recent electricity dispute resulted in power black-outs with unfortunate consequences for organizations relying on computers. Article discusses the implications of similar events in Britain in the future when computers are even more widely in use (1 1/2 pages).

  14. 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.)

  15. An Analysis of Loss of Offsite Power Sequence for the Severe Accident Analysis Database (II)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Dong Ha

    2006-12-15

    This report contains analysis methodologies and calculation results of loss of offsite 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, Twelve accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the loss of offsite 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 of loss of offsite sequence in this report will be utilized as input data of the severe accident analysis database system. This report updates and complements a previously published Technical Report.

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

    International Nuclear Information System (INIS)

    Clauss, D.B.; Wilson, R.K.; Blower, D.F.; Campbell, K.L.

    1994-06-01

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

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

    International Nuclear Information System (INIS)

    Hashimoto, Kazuichiro

    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.)

  18. An Analysis of Medium Loss of Coolant Sequence for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of medium loss of Coolant 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, 10 accident scenarios, which was predicted to have more than 10{sup -10} /ry occurrence frequency, have been analyzed as base cases for the medium 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 of the severe accident analysis database system.

  19. An Analysis of Large Loss of Coolant Sequence for the Severe Accident Analysis DB

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of Large loss of Coolant 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, 14 accident scenarios, which was predicted to have more than 10{sup -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 of the severe accident analysis database system.

  20. An Analysis of Small Loss of Coolant Sequence for the Severe Accident Analysis Database

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Song, Yong Mann

    2007-12-15

    This report contains analysis methodologies and calculation results of small loss of Coolant 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, 10 accident scenarios, which was predicted to have more than 10{sup -9} /ry occurrence frequency, have been analyzed as base cases for the small 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 of the severe accident analysis database system.

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  2. Assessment of severe accident source terms in pressurized-water reactors with a 40% mixed-oxide and 60% low-enriched uranium core using MELCOR 1.8.5.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.; Goldmann, Andrew S. (Texas A& M University, College Station, TX); Wagner, Kenneth C.; Powers, Dana Auburn; Ashbaugh, Scott G.; Longmire, Pamela

    2010-04-01

    As part of a Nuclear Regulatory Commission (NRC) research program to evaluate the impact of using mixed-oxide (MOX) fuel in commercial nuclear power plants, a study was undertaken to evaluate the impact of the usage of MOX fuel on the consequences of postulated severe accidents. A series of 23 severe accident calculations was performed using MELCOR 1.8.5 for a four-loop Westinghouse reactor with an ice condenser containment. The calculations covered five basic accident classes that were identified as the risk- and consequence-dominant accident sequences in plant-specific probabilistic risk assessments for the McGuire and Catawba nuclear plants, including station blackouts and loss-of-coolant accidents of various sizes, with both early and late containment failures. Ultimately, the results of these MELCOR simulations will be used to provide a supplement to the NRC's alternative source term described in NUREG-1465. Source term magnitude and timing results are presented consistent with the NUREG-1465 format. For each of the severe accident release phases (coolant release, gap release, in-vessel release, ex-vessel release, and late in-vessel release), source term timing information (onset of release and duration) is presented. For all release phases except for the coolant release phase, magnitudes are presented for each of the NUREG-1465 radionuclide groups. MELCOR results showed variation of noble metal releases between those typical of ruthenium (Ru) and those typical of molybdenum (Mo); therefore, results for the noble metals were presented for Ru and Mo separately. The collection of the source term results can be used as the basis to develop a representative source term (across all accident types) that will be the MOX supplement to NUREG-1465.

  3. Progress in core and fuel modelling to calculate severe accidents

    International Nuclear Information System (INIS)

    Bonnet, M.; Baldi, St.; Porta, J.

    2000-01-01

    The use of CERMET type composite fuels lead to a correct use of plutonium; a good thermomechanical behaviour due to a low operating temperature thanks to a high thermo-conductivity, that favours high burn-up due to the low fission gas release. However, the increase in the metallic mass, an alloy of zircaloy in the core, as well as the composite nature of the fuel with two very different melting temperatures (∼ 1,600 deg C for the metal, and 2,300 deg C for the ceramic) lead to a behaviour very different from that of the traditional ceramic fuel in the event of an accident. (authors)

  4. Integrated computer codes for nuclear power plant severe accident analysis

    International Nuclear Information System (INIS)

    Jordanov, I.; Khristov, Y.

    1995-01-01

    This overview contains a description of the Modular Accident Analysis Program (MAAP), ICARE computer code and Source Term Code Package (STCP). STCP is used to model TMLB sample problems for Zion Unit 1 and WWER-440/V-213 reactors. Comparison is made of STCP implementation on VAX and IBM systems. In order to improve accuracy, a double precision version of MARCH-3 component of STCP is created and the overall thermal hydraulics is modelled. Results of modelling the containment pressure, debris temperature, hydrogen mass are presented. 5 refs., 10 figs., 2 tabs

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-10-01

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

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

    International Nuclear Information System (INIS)

    Sairanen, R.

    1997-10-01

    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

  7. Evaluation of severe accident safety system value based on averting financial risks

    International Nuclear Information System (INIS)

    Hatch, S.W.; Benjamin, A.S.; Bennett, P.R.

    1983-01-01

    The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions

  8. Evaluation of severe accident safety system value based on averting financial risks. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Hatch, S.W.; Bennett, P.R.; Benjamin, A.S.

    1983-01-01

    The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions.

  9. Assessment of Equipment Capability to Perform Reliably under Severe Accident Conditions

    International Nuclear Information System (INIS)

    2017-07-01

    The experience from the last 40 years has shown that severe accidents can subject electrical and instrumentation and control (I&C) equipment to environmental conditions exceeding the equipment’s original design basis assumptions. Severe accident conditions can then cause rapid degradation or damage to various degrees up to complete failure of such equipment. This publication provides the technical basis to consider when assessing the capability of electrical and I&C equipment to perform reliably during a severe accident. It provides examples of calculation tools to determine the environmental parameters as well as examples and methods that Member States can apply to assess equipment reliability.

  10. Criteria for successful core reflood under severe accident conditions

    International Nuclear Information System (INIS)

    Hering, W.; Homann, Ch.

    2004-01-01

    In present German nuclear power plants, safety is enhanced by prescription of preemptive measures and accident mitigation measures. If preemptive measures (e.g. backfitting, improved safety procedures) should fail, mitigative procedures are foreseen to take credit of all safety-related systems available on site. As usual in daily life, no guarantee of complete success of accident management measures can be given, because the success of a core reflood depends essentially on the actual core state and its history, system pressure, and the injection rate of the activated reflood system. In the present work, available experimental data on core reflood are reviewed to define characteristic regimes and dependencies as well as identifying areas where experimental data are lacking, e.g. reflooding of large in-core debris/pool configurations. A reflood map is proposed based on core state and reflood mass flow rate. Common features of the behavior are deduced on the assumption that non-prototypic facility-based effects can be excluded. (authors)

  11. [Characterization of severe acute occupational poisoning accidents in China between 1989 and 2003].

    Science.gov (United States)

    Zhang, Min; Li, Tao; Wang, Huan-Qiang; Wang, Hong-Fei; Chen, Shu-Yang; Du, Xie-Yi; Zhang, Shuang; Qin, Jian

    2006-12-01

    To analyze severe acute occupational poisoning accidents reported in China between 1989 and 2003, and to study the characteristics of severe acute occupational poisoning accidents and provide scientific evidences for prevention and control strategies. The data from the national occupational poisoning case reporting system were analyzed with descriptive methods. (1) There were 506 acute severe occupational poisoning accidents for 15 years with 4 657 workers poisoned. The total poisoning rate was 54.8%, and the total mortality was 16.5%. The average poisoning age was (31.9 +/- 9.8) years old and the average death age was (33.7 +/- 10.3) years old. The poisoning accidents occurred more in men than in women. (2) There were more than 112 chemicals which caused these poisoning accidents. Most of the accidents caused by hydrogen sulfide, carbon monoxide, benzene and homologs, metal and metalloid and carbon dioxide, and the types of chemicals varied in different types of industries. (3) The accidents mainly occurred in chemical industry, manufacture, water disposal industry, mining and construction industry, and the risk was higher in some jobs than others, such as cleanout, machine maintenance and repair, production, mine and digging. The accidents occurred more frequently from April to August each year. (1) The control over the severe acute occupational poisoning is urgent. (2) The trend of the characteristics of severe acute occupational poisoning accidents is centralized in the high risk industries, poisons and jobs. (3) The characteristics of the accidents varied in different types of industries. (4) It is the key point to strengthen the supervision on poisoning.

  12. Example of severe accident management guidelines validation and verification using full scope simulator

    International Nuclear Information System (INIS)

    Krajnc, B.; Basic, I.; Spiler, J.

    2001-01-01

    The purpose of Severe Accident Management Guidelines (SAMG) is to provide guidelines to mitigate and control beyond design bases accidents. These guidelines are to be used by the technical support center that is established at the plant within one hour after the beginning of the accident as a technical support for the main control room operators. Since some of the accidents can progress very fast there are also two guidelines provided for the main control room operators. The first one is to be used if the core damage occurs and the TSC is not established yet and the second one after technical support center become operational. After SG replacement and power uprate in year 2000, NPP Krsko developed Rev.1 of these procedures, which have been validated and verified during one-week effort. Plant specific simulator capable of simulating severe accidents was extensively used.(author)

  13. 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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. Mitigation of severe accidents in light water reactors: Chapter 8

    International Nuclear Information System (INIS)

    Kastenberg, W.E.; Catton, I.

    1983-01-01

    As part of the NRC program on degraded core and core-melt accidents beyond the design basis, the work presented here focuses on containment mitigation systems. Included are studies aimed at estimating the risk reduction potential for filtered-vented containment systems, passive containment heat removal systems, and features to mitigate against hydrogen burns and base mat penetration. Specific aspects of mitigation for Zion, Indian Poin and Limerick plants are considered. For Zion, consideration of a filtered-vented containment system and a passive containment heat removal system was considered. For Indian Point, the use of heat pipes for passive heat removal was considered. Lastly, for Limerick a low-volume filtered venting system was found to provide a risk reduction factor on the order of 17, when based on man-rem reduction

  16. Quality improvements of thermodynamic data applied to corium interactions for severe accident modelling in SARNET2

    Czech Academy of Sciences Publication Activity Database

    Bakardjieva, Snejana; Barrachin, M.; Bechta, S.; Bezdička, Petr; Bottomley, D.; Brissoneau, L.; Cheynet, B.; Dugne, O.; Fischer, E.; Fischer, M.; Gusarov, V.; Journeau, C.; Khabensky, V.; Kiselová, M.; Manara, D.; Piluso, P.; Sheindlin, M.; Tyrpekl, V.; Wiss, T.

    2014-01-01

    Roč. 74, SI (2014), s. 110-124 ISSN 0306-4549 Institutional support: RVO:61388980 Keywords : Corium * Severe accidents * Thermodynamic database Subject RIV: CA - Inorganic Chemistry Impact factor: 0.960, year: 2014

  17. 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.

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

    International Nuclear Information System (INIS)

    Yoo, Seung J.; Gu, Beom W.; Nguyen, Duy T.; Choi, Bo H.; Rim, Chun T.; Lee, So I.

    2014-01-01

    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

  19. Sisifo-gas a computerised system to support severe accident training and management

    International Nuclear Information System (INIS)

    Castro, A.; Buedo, J.L.; Borondo, L.; Lopez, N.

    2001-01-01

    Nuclear Power Plants (NPP) will have to be prepared to face the management of severe accidents, through the development of Severe Accident Guides and sophisticated systems of calculation, as a supporting to the decision-making. SISIFO-GAS is a flexible computerized tool, both for the supporting to accident management and for education and training in severe accident. It is an interactive system, a visual and an easily handle one, and needs no specific knowledge in MAAP code to make complicate simulations in conditions of severe accident. The system is configured and adjusted to work in a BWR/6 technology plant with Mark III Containment, as it is Cofrentes NPP. But it is easily portable to every other kind of reactor, having the level 2 PSA (probabilistic safety analysis) of the plant to be able to establish the categories of the source term and the most important sequences in the progression of the accident. The graphic interface allows following in a very intuitive and formative way the evolution and the most relevant events in the accident, in the both system's way of work, training and management. (authors)

  20. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    International Nuclear Information System (INIS)

    Hodge, S.A.; Kress, T.S.; Cleveland, J.C.; Petek, M.

    1992-01-01

    This paper briefly describes the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to evaluate the effectiveness and feasibility of current and proposed strategies for BWR severe accident management. These results are described in detail in the just-released report Identification and Assessment of BWR In-Vessel Severe Accident Mitigation Strategies, NUREG/CR-5869, which comprises three categories of findings. First, an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences is combined with a review of the BWR Owners' Group Emergency Procedure Guidelines (EPGs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, two of the four candidate strategies identified by this effort are assessed in detail. These are (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored

  1. Proposal strategy and policy on nuclear safety for no-more severe accidents

    International Nuclear Information System (INIS)

    2013-01-01

    Following the outspoken advice saying 'scientists and engineers concerning with nuclear power promotion and safety should be responsible for clarifying how preventable or what measures should be needed to prevent severe accidents occurring at Fukushima Daiichi nuclear power plants (NPPs)', committee on prevention of severe accidents at NPPs was established by relevant nuclear scientists and engineers involved so as to discuss basic issues to be solved from scientific and technical viewpoints. Based on the review of 'defense in depth' concept and accident analysis at Fukushima nuclear accident, four major proposals and six supplements to be established were identified such as: (1) finding mechanism of beyond imagination events for natural disaster, terrorism, and internal events, (2) reform of comprehensive safety standards and guidelines with performance basis easy to reflect latest knowledge and technology as 'back-fitting', (3) severe accidents measures, their validation, and drilling on accident management to advance procedures and develop human resources, and (4) risk communications and public disclosure of information. This article described backgrounds of committee's proposals on nuclear safety for no-more severe accidents. (T. Tanaka)

  2. Safety upgrading activities against tsunami, earthquake, and severe accident at Hamaoka NPPs

    International Nuclear Information System (INIS)

    Watanabe, Tetsuya; Wakunaga, Takao; Ishida, Takahisa

    2013-01-01

    As the lessons learned by the Fukushima Daiichi NPPs accident, Chubu Electric Power carried out the Emergency Safety Measures at Hamaoka NPPs immediately, and announced the plan for tsunami countermeasures including the construction of 18m-height tsunami protection wall in July 2011. Furthermore, the company announced the additional severe accident and tsunami countermeasures, and etc. in December 2012 and in April 2013, such as the installation of Filtered Containment Venting System and increasing the height of the tsunami protection wall from 18m to 22m. In this paper, we present major safety upgrading activities against tsunami, earthquake and severe accident at Hamaoka NPPs. (author)

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

    International Nuclear Information System (INIS)

    Yvon, M.; Lohnert, G.; Lauret, P.; Bittermann, D.

    1998-01-01

    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

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

    International Nuclear Information System (INIS)

    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

  5. A systematic framework for effective uncertainty assessment of severe accident calculations; Hybrid qualitative and quantitative methodology

    International Nuclear Information System (INIS)

    Hoseyni, Seyed Mohsen; Pourgol-Mohammad, Mohammad; Tehranifard, Ali Abbaspour; Yousefpour, Faramarz

    2014-01-01

    This paper describes a systematic framework for characterizing important phenomena and quantifying the degree of contribution of each parameter to the output in severe accident uncertainty assessment. The proposed methodology comprises qualitative as well as quantitative phases. The qualitative part so called Modified PIRT, being a robust process of PIRT for more precise quantification of uncertainties, is a two step process for identifying and ranking based on uncertainty importance in severe accident phenomena. In this process identified severe accident phenomena are ranked according to their effect on the figure of merit and their level of knowledge. Analytical Hierarchical Process (AHP) serves here as a systematic approach for severe accident phenomena ranking. Formal uncertainty importance technique is used to estimate the degree of credibility of the severe accident model(s) used to represent the important phenomena. The methodology uses subjective justification by evaluating available information and data from experiments, and code predictions for this step. The quantitative part utilizes uncertainty importance measures for the quantification of the effect of each input parameter to the output uncertainty. A response surface fitting approach is proposed for estimating associated uncertainties with less calculation cost. The quantitative results are used to plan in reducing epistemic uncertainty in the output variable(s). The application of the proposed methodology is demonstrated for the ACRR MP-2 severe accident test facility. - Highlights: • A two stage framework for severe accident uncertainty analysis is proposed. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • Uncertainty importance measure quantitatively calculates effect of each uncertainty source. • Methodology is applied successfully on ACRR MP-2 severe accident test facility

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

    OpenAIRE

    Dumrak, Jantanee; Mostafa, Sherif; Kamardeen, Imriyas; Rameezdeen, Raufdeen

    2013-01-01

    While the causes of accidents in the construction industry have been extensively studied, severity remains an understudied area. In order to provide more evidence for the currently limited number of empirical investigations on severity, this study analysed 24,764 construction accidents reported during 2002-11 in South Australia. A conceptual model developed through literature uses personal characteristics such as age, experience, gender and language. It also employs work-related factors such ...

  7. Severe Accident Mitigation by using Core Catcher applicable for Korea standard nuclear power plant

    International Nuclear Information System (INIS)

    Park, Hae Kyun; Kim, Sang Nyung

    2013-01-01

    Nuclear power plants have been designed and operated in order to prevent severe accident because of their risk that contains tremendous radioactive materials that are potentially hazardous. Moreover, the government requested the nuclear industry to implement a severe accident management strategy for existing reactors to mitigate the risk of potential severe accidents. However, Korea standard nuclear power plant(APR-1400 and OPR-1000) are much more vulnerable for severe accident management than that of developed countries. Due to the design feature of reactor cavity in Korea standard nuclear power plant, inequable and serious Molten Core-Concrete Interaction(MCCI) may cause considerable safety problem to the reactor containment liner. At worst, it brings the release of radioactive materials to the environment. This accident applies to the fourth level of defense in depth(IAEA 1996), 'severe accident'. This study proposes and designs the 'slope' to secure reactor containment liner integrity when the corium spreads out from the destroyed reactor vessel to the reactor cavity due to the core melting accident. For this, make the initial corium distribution evenly exploit the 'slope' on the basis of the study of Ex-vessel corium behavior to prevent inequable and serious MCCI, in order to mitigate severe accident. The viscosity has a dominant position in the calculation. According to the result, the spread out distance on the slope is 10.7146841m, considering the rough surface of the concrete(slope) and margin of reactor cavity end(under 11m). Easy to design, production and economic feasibility are the advantage of the designed slope in this study. However, the slope design may unsuitable when the sequences of the accidents did not satisfy the assumptions as mentioned. Despite of those disadvantages, the slope will show a great performance to mitigate the severe accident. As mentioned in assumption, the corium releasing time property was conservatively calculated

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

    International Nuclear Information System (INIS)

    Silva, Kampanart; Promping, Jiraporn; Okamoto, Koji; Ishiwatari, Yuki

    2014-01-01

    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)

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

    International Nuclear Information System (INIS)

    McKenna, T.J.; Giitter, J.

    1987-01-01

    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)

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

    International Nuclear Information System (INIS)

    McKenna, T.J.; Glitter, J.G.

    1988-10-01

    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

  11. Blackout cloth for dormancy induction

    Science.gov (United States)

    Tom Jopson

    2007-01-01

    The use of blackout cloth to create long night photoperiods for the induction of dormancy in certain conifer species has been an established practice for a long time. Its use was suggested by Tinus and McDonald (1979) as an effective technique, and the practice has been commonly used in Canadian forest nurseries for a number of years. Cal-Forest Nursery installed its...

  12. Identification and assessment of BWR in-vessel severe accident mitigation strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.; Cleveland, J.C.; Kress, T.S.; Petek, M. [Oak Ridge National Lab., TN (United States)

    1992-10-01

    This report provides the results of work carried out in support of the US Nuclear Regulatory Commission Accident Management Research Program to develop a technical basis for evaluating the effectiveness and feasibility of current and proposed strategies for boiling water reactor (BWR) severe accident management. First, the findings of an assessment of the current status of accident management strategies for the mitigation of in-vessel events for BWR severe accident sequences are described. This includes a review of the BWR Owners` Group Emergency Procedure Guidelines (EPGSs) to determine the extent to which they currently address the characteristic events of an unmitigated severe accident and to provide the basis for recommendations for enhancement of accident management procedures. Second, where considered necessary, new candidate accident management strategies are proposed for mitigation of the late-phase (after core damage has occurred) events. Finally, recommendations are made for consideration of additional strategies where warranted, and two of the four candidate strategies identified by this effort are assessed in detail: (1) preparation of a boron solution for reactor vessel refill should control blade damage occur during a period of temporary core dryout and (2) containment flooding to maintain the core debris within the reactor vessel if the injection systems cannot be restored.

  13. Modeling the chemical behavior of radionuclides in severe nuclear accidents

    International Nuclear Information System (INIS)

    Gieseke, J.A.; Alexander, C.A.

    1986-01-01

    The chemical behavior of radionuclides can affect their eventual release to the environment during transport under accident conditions. Four areas or processes of importance are: release from the fuel and core region, reactions with gaseous environments and surfaces in the RCS, core-concrete interactions, and reactions with gases, surfaces and liquids in the containment. In the fuel itself, for example, the chemical potential of oxygen determines which complexes such as Cs 2 UO 4 may be formed in the solid state. Interactions between fission products and various gases, vapors, or surfaces (including deposits) may affect whether significant retention occurs in the RCS. For cases where molten components of the core may fall through to interact with the concrete cavity floor, the chemical composition of the melt and concrete and their chemical interactions are predicted to determine, to a large extent, the rates at which fission products are evolved. Finally, in the containment building atmosphere, the presence of volatile iodine species may result from chemical reactions. Various computer codes are available which provide predictions of chemical behavior in these four areas of importance. The results demonstrate the important role chemistry can have in affecting release of radionuclides to the environment

  14. Release fraction of PWR after severe accidents. Vol. 4

    International Nuclear Information System (INIS)

    Aziz, M.; El-Messeiry, A.M.

    1996-01-01

    Fission fragments and gases are emitted after accidents as a result of core meltdown and core concrete interactions. These aerosols are transported and fill the reactor containment. With increasing the pressure above pressure design bases, a failure of containment may occur and subsequently these aerosols will release into the external environment leading to a source term of radioactivity that affects the safety of workers and public. The amount of aerosol which escapes to the environment can be described by the release fraction which is defined as the total accumulated aerosol which initially enters the containment. The factors that affect the release fraction is studied, and the aerosol dynamics equation is used to model the release of aerosol to the outside atmosphere. These factors are containment pressure, failure time,break area, the size of aerosol particle. It found that early failure time and higher pressure increase the release fraction, also the release faction is affected by the area and the aerosol particle size. 7 figs., 2 tabs

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

    International Nuclear Information System (INIS)

    Kolev, N.I.

    2001-01-01

    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)

  16. 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.

  17. System 80+TM PRA insights on severe accident prevention and mitigation

    International Nuclear Information System (INIS)

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

    2004-01-01

    The System 80 + design is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the ALWR Utility Requirements Document (URD), and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the System 80 + design are described. The results of the System 80 + PRA are presented and the insights gained from the PRA sensitivity analyses are discussed. ABB-CE considered defense-in-depth for accident prevention and mitigation early in the design process and used robust design features to ensure that the System 80 + design achieved a low core damage frequency, low containment conditional failure probability, and excellent deterministic containment performance under severe accident conditions and to ensure that the risk was properly allocated among design features and between prevention and mitigation. (author)

  18. 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 (p<0.05). These findings can be applied in the design and implementation of a comprehensive safety and health risk management system to reduce ASR.

  19. 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)

  20. Comparative study of the hydrogen generation during short term station blackout (STSBO) in a BWR

    International Nuclear Information System (INIS)

    Polo-Labarrios, M.A.; Espinosa-Paredes, G.

    2015-01-01

    Highlights: • Comparative study of generation in a simulated STSBO severe accident. • MELCOR and SCDAP/RELAP5 codes were used to understanding the main phenomena. • Both codes present similar thermal-hydraulic behavior for pressure and boil off. • SCDAP/RELAP5 predicts 15.8% lower hydrogen production than MELCOR. - Abstract: The aim of this work is the comparative study of hydrogen generation and the associated parameters in a simulated severe accident of a short-term station blackout (STSBO) in a typical BWR-5 with Mark-II containment. MELCOR (v.1.8.6) and SCDAP/RELAP5 (Mod.3.4) codes were used to understand the main phenomena in the STSBO event through the results comparison obtained from simulations with these codes. Due that the simulation scope of SCDAP/RELAP5 is limited to failure of the vessel pressure boundary, the comparison was focused on in-vessel severe accident phenomena; with a special interest in the vessel pressure, boil of cooling, core temperature, and hydrogen generation. The results show that at the beginning of the scenario, both codes present similar thermal-hydraulic behavior for pressure and boil off of cooling, but during the relocation, the pressure and boil off, present differences in timing and order of magnitude. Both codes predict in similar time the beginning of melting material drop to the lower head. As far as the hydrogen production rate, SCDAP/RELAP5 predicts 15.8% lower production than MELCOR

  1. Analysis of severe accidents on fast reactor test loop

    International Nuclear Information System (INIS)

    Cenerini, R.; Verzelletti, G.; Curioni, S.

    1975-01-01

    The Pec reactor is a sodium cooled fast reactor which is being designed for the primary purpose of accomodating closed sodium cooled test loops for the developmental and proof testing of fast reactor fuel assemblies. The test loops are located in the central test region of reactor. The basic function for which the loop is designed is burn-up to failure testing of fuel under advanced performance conditions. It is therefore necessary to design the loop for failure conditions. Basically two types of accidents can occur within the loops: rupture of gas plenum in the fuel pins and coolant starvation. Explosive tests on Pec loop, whose first set is described in this report, are devoted to investigate the effects of an accidental energy release on loop containment. The loop model reproduces in the test section the prototype dimensions in radial scale 1:1. Using a wire explosive charge of 300mm, the height of test section is sufficient for determining the containment capability of the loop that has a nearly constant deformation in a length of. 3-4 time the diameter. The inertial effects of the coolant column are reproduced by two tubes at the extremities of test section, closed with top plugs. Some tests has been performed by wrapping around the test section four layers of steel wire in order to evaluate the influence on the containment of tungsten wire that is foreseen in prototype loop. The influence of the coolant around the loop was evaluated by inserting the model in water. Dummy sub-assemblies was used and explosive substitutes the central rods. Piezoelectric pressure transducers were mounted on the three plugs and radial deformation was measured directly at different height. From experiments performed it resulted the importance of harmonic wires and inertial reaction of external water on loop containment; maximum containable energy is about 50 Cal with E.1 explosive

  2. Blackout sequence modeling for Atucha-I with MARCH3 code

    International Nuclear Information System (INIS)

    Baron, J.; Bastianelli, B.

    1997-01-01

    The modeling of a blackout sequence in Atucha I nuclear power plant is presented in this paper, as a preliminary phase for a level II probabilistic safety assessment. Such sequence is analyzed with the code MARCH3 from STCP (Source Term Code Package), based on a specific model developed for Atucha, that takes into accounts it peculiarities. The analysis includes all the severe accident phases, from the initial transient (loss of heat sink), loss of coolant through the safety valves, core uncovered, heatup, metal-water reaction, melting and relocation, heatup and failure of the pressure vessel, core-concrete interaction in the reactor cavity, heatup and failure of the containment building (multi-compartmented) due to quasi-static overpressurization. The results obtained permit to visualize the time sequence of these events, as well as provide the basis for source term studies. (author) [es

  3. Modelling and forecasting occupational accidents of different severity levels in Spain

    International Nuclear Information System (INIS)

    Carmen Carnero, Maria; Jose Pedregal, Diego

    2010-01-01

    The control of accidents at the work place is a critical issue all over the world. The consequences of occupational accidents in terms of costs for the company in which the accidents take place is only one minor matter, being the social impact and the loss of human life the most controversial effects of this important problem. The methods used to forecast future evolution of accidents are often limited to trend estimations and projections, being the scientific literature on this topic rather scarce. This paper aims at showing and predicting the evolution of Spanish occupational accidents of different levels of severity, allowing the evaluation of the influence that preventive actions carried out by public administrations or private companies may have over the number of occupational accidents. Though some contributions may be found on this topic for Spain, this paper is the first contribution that forecast occupational accidents for different levels of severity using Multivariate Unobserved Components models developed in a State Space framework extended to deal with the irregular sampling interval of the data. Data from 1998 to 2009 have been used to test the efficacy of the forecasting system.

  4. Including severe accidents in the design basis of nuclear power plants: An organizational factors perspective after the Fukushima accident

    International Nuclear Information System (INIS)

    Alvarenga, M.A.B.; Frutuoso e Melo, P.F.

    2015-01-01

    Highlights: • The Fukushima accident was man-made and not caused by natural phenomena. • Vulnerabilities were known by regulator and licensee but measures were not taken. • There was lack of independence and transparency of the regulatory body. • Laws and regulations have not been updated to international standards. • Organizational failures have played an important role in the Fukushima accident. - Abstract: The Fukushima accident was clearly an accident made by humans and not caused by natural phenomena as was initially thought. Vulnerabilities were known by both regulators and operator but they postponed measures. The emergency plan was not effective in protecting the public, because the involved parties were not sufficiently prepared to make the right decisions. The shortcomings and faults mentioned above resulted from the lack of independence and transparency of the regulatory body. Even laws and regulations, and technical standards, have not been upgraded to international standards. Regulators have not defined requirements and left for the operator to decide what would be more appropriate. In this aspect, there was clearly a lack of independence between these bodies and operator’s lobby power. The above situation raised the question of urgent updating of institutions, in particular those responsible for nuclear safety. The above evidences show that several nuclear safety principles were not followed. This paper intends to highlight some existing safety criteria that were developed from the operational experience of the severe accidents that occurred at TMI and Chernobyl that should be incorporated in the design of new nuclear power plants and to provide appropriate design changes (backfittings) for reactors that belong to the previous generation prior to the occurrence of these accidents, through the study of design vulnerabilities. Furthermore, the main criteria that define an effective regulatory agency are also discussed. Although these

  5. Validation Method of a Telecommunications Blackout Attack

    National Research Council Canada - National Science Library

    Amado, Joao; Nunes, Paulo

    2005-01-01

    This paper presents an evaluation method of telecommunications infrastructure vulnerabilities, allowing the identification of components that can be attacked in order to achieve a communications blackout...

  6. Second Specialist Meeting on operator aids for severe accident management: summary and conclusions

    International Nuclear Information System (INIS)

    1997-01-01

    The second OECD Specialist Meeting on operator aids for severe accident management (SAMOA-2) was held in Lyon, France (1997), and was attended by 33 specialists representing ten OECD member countries. As for SAMOA-1, 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. 20 papers were presented; there were two demonstrations of computerized systems (the ATLAS analysis simulator developed by GRS, and EDF's 'Simulateur Post Accidentels' (SIPA). The three sessions dealt with operator aids for control rooms, operator aids for technical support centres, and simulation tools for operator training. The various papers for each session are summarized

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

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

    International Nuclear Information System (INIS)

    1992-01-01

    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.

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

    International Nuclear Information System (INIS)

    Viktorov, A.

    2015-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-12-15

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

  11. Development of Severe Accident Management Strategies for Shin-Kori 3 and 4

    International Nuclear Information System (INIS)

    Lee, Youngseung; Kim, Hyeongtaek; Shin, Jungmin

    2013-01-01

    Shin-Kori units 3 and 4 are new reactors under construction as an APR 1400 type reactor. The plants which considered coping with severe accident from design phase are different from other operating plants in view of severe accident management strategies. The purpose of this paper is to establish optimal strategies for Shin-Kori 3 and 4. A scheme for optimized severe accident management was drawn up with the object of achieving core cooling, containment integrity, and decreased release of fission product. Shin-Kori units 3 and 4 are a new reactor and designed to add mitigating systems for coping with severe accident such as ECSBS, PAR, and CFS. Also the plants are reflected as a part of Fukushima followup measures The strategies of SAMG for Shin-Kori 3 and 4 were developed. The strategic approach was based on the concept of defense in depth. Firstly, strategies for core cooling were chosen such as RCS depressurization, injection to SG, injection to RCS, and injection to reactor cavity. Secondly, the plans for containment integrity were developed for controlling pressure and hydrogen in containment. Lastly, reduced release of fission product was considered for protection of the public after containment failure. The achieved strategies meet the needs of effective methods for severe accident management and enhancement of safety

  12. Radiological consequences in Denmark from a severe reactor accident in the Ignalina power plant

    International Nuclear Information System (INIS)

    Lauritzen, B.; Damkjaer, A.; Nielsen, F.; Nielsen, S.P.; Nonboel, E.; Thykier-Nielsen, S.

    1996-05-01

    The radiological consequences in Denmark are assessed, following a hypothetical severe reactor accident in the Lithuanian nuclear power plant at Ignalina. The release of radionuclides and subsequent spreading in the atmosphere is initially assumed to be similar in magnitude to that of the Chernobyl accident in April 1986. The atmospheric transport and deposition of radionuclides from the Ignalina power plant is modelled as a stochastic process, and a probability distribution is estimated for the deposition on Danish territory, based on the deposition pattern of radiocaesium after the Chernobyl accident. At the 1% probability level of the atmospheric transport and deposition, the collective radiation dose to the Danish population amounts to 100,000 manSv, assuming the accident to happen in the summer months shortly before harvest. The most important pathway for radiation dose to the population will be ingestion of contaminated food,and restrictions on foodstuff are likely to be required. (au) 15 tabs., 7 ills., 28 refs

  13. Alcohol-induced blackout as a criminal defense or mitigating factor: an evidence-based review and admissibility as scientific evidence.

    Science.gov (United States)

    Pressman, Mark R; Caudill, David S

    2013-07-01

    Alcohol-related amnesia--alcohol blackout--is a common claim of criminal defendants. The generally held belief is that during an alcohol blackout, other cognitive functioning is severely impaired or absent. The presentation of alcohol blackout as scientific evidence in court requires that the science meets legal reliability standards (Frye, FRE702/Daubert). To determine whether "alcohol blackout" meets these standards, an evidence-based analysis of published scientific studies was conducted. A total of 26 empirical studies were identified including nine in which an alcohol blackout was induced and directly observed. No objective or scientific method to verify the presence of an alcoholic blackout while it is occurring or to confirm its presence retrospectively was identified. Only short-term memory is impaired and other cognitive functions--planning, attention, and social skills--are not impaired. Alcoholic blackouts would not appear to meet standards for scientific evidence and should not be admissible. © 2013 American Academy of Forensic Sciences.

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

    International Nuclear Information System (INIS)

    2002-01-01

    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

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

    International Nuclear Information System (INIS)

    Okamoto, Koji; Sugiyama, Tomoyuki; Kamata, Shinya

    2014-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Rameez Rameezdeen

    2017-01-01

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

  17. Response to severe-accident policy statement: Boiling water reactor containment vulnerability assessment

    International Nuclear Information System (INIS)

    Gabor, J.R.; Burns, E.T.; Mairs, T.P.

    1989-01-01

    The U.S. Nuclear Regulatory Commission (NRC's) Severe-Accident Policy Statement, Safety Goal Policy Statement, Individual Plant Examination (IPE) Generic Letter, and Containment Performance Improvement (CPI) Program seek to characterize adequate containment performance. This paper describes a framework (developed through the cooperation of a number of utilities) within which the following can be accomplished: questions regarding plant-specific containment performance can be addressed; the impact of proposed plant modifications can be investigated and the results communicated to the NRC. The NRC is currently assessing the performance of all containment types under postulated severe-accident conditions. Issues have been raised by the NRC regarding containment performance because it is a final barrier protecting the public against the release of radionuclides under sever-accident conditions. In addition, there are several arenas where additional related issues may be raised, e.g., NUREG-1150 (final issue), IPE reviews by the NRC staff, recommended accident management strategies, accident management proposed generic letter, and the NRC generic evaluation of boiling water reactor. This paper presents the methodology developed in cooperation with a number of utilities to respond to the NRC initiatives requiring a plant-specific containment performance evaluation as part of the IPE process

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

    Science.gov (United States)

    Rameezdeen, Rameez; Elmualim, Abbas

    2017-01-11

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

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

    International Nuclear Information System (INIS)

    Sonnenkalb, Martin

    1992-07-01

    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

  20. Zircaloy oxidation and hydrogen generation behavior during severe accidents

    International Nuclear Information System (INIS)

    Cronenberg, A.W.; Miller, R.W.; Osetek, D.J.

    1987-01-01

    Zircaloy oxidation and H 2 generation data are presented for recent in-pile severe fuel damage tests. The principal questions investigated concern zircaloy melting and bundle reconfiguration effects on oxidation behavior. A comparison of H 2 generation and cladding temperature data indicate that significant oxidation occurred after the onset of fuel liquefaction. Posttest metallographic observations of the test debris also indicate a high degree of oxidation of once-molten zircaloy. Analysis of bundle reconfiguration effects indicate that essentially complete flow area blockage (>98%) would be required to retard steam flow through the degraded bundle so as to diminish H 2 production. Such extreme blockage conditions are not supported by posttest bundle examination

  1. Preliminary Assessment of Depressurization Performance of Reactor Building Spray dedicated to Severe Accident

    International Nuclear Information System (INIS)

    Park, Chang Hwan; Jo, Jae Sun; Yoon, Sun Hong

    2010-01-01

    In these days, the global demand for the nuclear power plant is gradually increasing and then it is encouraging to see the mood in which the possibility of exportation of Korean has been realized. According to this situation, the need for development of the country-tailored NPP is emerging because that there are some differences among the safety requirements of each country. Especially, European countries require relatively conservative safety criteria for the severe accident. Thus, development of a tactical NPP with the enhanced safety features dedicated to the severe accident is on the way. One of these safety features is the containment spray system dedicated to the severe accident. In this study, the depressurization capacity of the SA spray is assessed and the minimum capacity ensuring applicable performance is estimated with MAAP4 code. The reference plant for this analysis is chosen as APR1400

  2. Annual technical meeting of the NRC cooperative severe accident research program

    International Nuclear Information System (INIS)

    Silver, E.G.

    1993-01-01

    This brief report summarizes the 1992 annual technical meeting of the NRC Cooperative Severe Accident Research Program (CSARP-92) held at the Hyatt Regency Hotel in Bethesda, Maryland, May 4-8, 1992. The report is taken mainly from coverage of the meeting published in the June 5, 1992, issue of Atomic Energy Clearinghouse. Results of this meeting are formalized at the Water Reactor Safety Information Meetings (WRSIM) that are held annually in October. Nuclear Safety summarizes the annual WRSIM meetings and provides a list of the presentations that were given. Interested readers are encouraged to review listed topics to identify specific topic areas in severe accident research. Sessions were held on in-vessel core melt progression; fuel-coolant interactions; fission-product behavior; direct containment heating; and severe accident code development, assessment, and validation. Summaries of the individual technical sessions and the current state of the art in these areas were given by the chairmen

  3. Evaluation of the Layer Inversion of Melt Pool during the Severe Accident in the APR1400

    International Nuclear Information System (INIS)

    Kang, Kyoung-Ho; Park, Rae-Joon; Hong, Seong-Wan

    2008-01-01

    During the severe accidents, thermal load from the oxidic pool can be concentrated on the side wall of the RPV due to the thermal barrier effect in the thin metallic layer. The focusing effect of the metallic layer is mainly determined by the molten pool configuration in the lower head of the RPV. Therefore, for the precise evaluations on the coolability through the in-vessel retention of corium during the severe accident, the melt pool configuration should be accurately defined. The melt pool configurations inside the lower head of the reactor vessel affect the initial thermal load to the vessel and play a key role in determining the integrity of the reactor vessel. In this study, thermodynamic analyses were performed to examine the final melt pool configuration during the severe accidents in the APR1400. As the representative accident scenarios, Large Break Loss of Coolant Accident (LBLOCA), Medium Break Loss of Coolant Accident (MBLOCA), Station Black Out (SBO), and Total Loss of Feed Water (TLFW) were considered. The initial melt pool conditions, such as melt mass and melt pool temperature etc., were calculated using the SCDAP/RELAP5/MOD3.3 code for each accident scenario of the APR1400. The thermodynamic analyses were performed using the GEMINI code. Combined with the GEMINI code calculations and the peer review on the RASPLAV/ MASCA experimental results, the final melt pool configuration in case of MBLOCA sequence was determined as a first step. Based on the thermodynamic analyses for the melt pool compositions, the possibility of the layer inversion between the oxidic pool and the metallic layer was examined

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  5. The Impact of Severe Nuclear Accidents on National Decision for Nuclear Decommissioning

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Young A; Hornibrook, Carol; Yim, Man Sung [KAIST, Daejeon (Korea, Republic of)

    2016-05-15

    Many researchers have tried to identify the impact of severe nuclear accidents on a country's or international nuclear energy policy [2-3]. However, there is little research on the influence of nuclear accidents and historical events on a country's decision to permanently shutdown an NPP versus international nuclear decommissioning trends. To demonstrate the correlation between a nuclear severe accident and the impact on world nuclear decommissioning, this research reviewed case studies of individual historical events, such as the St. Lucens, TMI, Chernobyl, Fukushima accidents and the series of events leading up to the collapse of the Soviet Union. For validation of the results of these case studies, a statistical analysis was conducted using the R code. This will be useful in explaining how international and national decommissioning strategies are affected by shutdown reasons, i.e. world historical events. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently In conclusion, nuclear severe accidents and historical events have an impact on the number of international NPPs that shutdown permanently and cancelled NPP construction. This directly impacts international nuclear decommissioning policy and nuclear energy policy trends. The number of permanently shutdown NPPs was selected as an indicator because any relationship between the number of permanently.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-08-01

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

  7. Severe accident management instrumentation in the Finnish NPP's

    Energy Technology Data Exchange (ETDEWEB)

    Tuomisto, H. (Imatran Voima Oy, Vantaa (Finland)); Felin, A. (Imatran Voima Oy, Vantaa (Finland)); Lucander, A. (Teollisuuden Voima Oy, Olkiluoto (Finland)); Tuuri, J. (Teollisuuden Voima Oy, Olkiluoto (Finland)); Koski, S. (Teollisuuden Voima Oy, Olkiluoto (Finland))

    1992-07-01

    The developmental stage of the severe accident management program of the Loviisa plant (VVER-440) has recently allowed the definition of instrumentation needs. The paper is aimed at discussing the principal approaches, how the plant-specific instrumentation needs have been derived from the safety functions of the severe accident management in each case. A distinction is made between the instrumentation that is of crucial importance for performing a correct management measure and the instrumentation needed for monitoring the success. New instrumentation is rather strictly limited to those ensuring the safety functions. (orig.)

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

    International Nuclear Information System (INIS)

    Karwat, H.

    1992-01-01

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

  9. Code package {open_quotes}SVECHA{close_quotes}: Modeling of core degradation phenomena at severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Veshchunov, M.S.; Kisselev, A.E.; Palagin, A.V. [Nuclear Safety Institute, Moscow (Russian Federation)] [and others

    1995-09-01

    The code package SVECHA for the modeling of in-vessel core degradation (CD) phenomena in severe accidents is being developed in the Nuclear Safety Institute, Russian Academy of Science (NSI RAS). The code package presents a detailed mechanistic description of the phenomenology of severe accidents in a reactor core. The modules of the package were developed and validated on separate effect test data. These modules were then successfully implemented in the ICARE2 code and validated against a wide range of integral tests. Validation results have shown good agreement with separate effect tests data and with the integral tests CORA-W1/W2, CORA-13, PHEBUS-B9+.

  10. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions - final report. Vol. 1

    International Nuclear Information System (INIS)

    Rogers, J.T.

    1984-06-01

    This report gives the results of a study of the thermo-hydraulic aspects of severe accident sequences in CANDU reactors. The accident sequences considered are the loss of the moderator cooling system and the loss of the moderator heat sink, each following a large loss-of-coolant accident accompanied by loss of emergency coolant injection. Factors considered include expulsion and boil-off of the moderator, uncovery, overheating and disintegration of the fuel channels, quenching of channel debris, re-heating of channel debris following complete moderator expulsion, formation and possible boiling of a molten pool of core debris and the effectiveness of the cooling of the calandria wall by the shield tank water during the accident sequences. The effects of these accident sequences on the reactor containment are also considered. Results show that there would be no gross melting of fuel during moderator expulsion from the calandria, and for a considerable time thereafter, as quenched core debris re-heats. Core melting would not begin until about 135 minutes after accident initiation in a loss of the moderator cooling system and until about 30 minutes in a loss of the moderator heat sink. Eventually, a pool of molten material would form in the bottom of the calandria, which may or may not boil, depending on property values. In all cases, the molten core would be contained within the calandria, as long as the shield tank water cooling system remains operational. Finally, in the period from 8 to 50 hours after the initiation of the accident, the molten core would re-solidify within the calandria. There would be no consequent damage to containment resulting from these accident sequences, nor would there be a significant increase in fission product releases from containment above those that would otherwise occur in a dual failure LOCA plus LOECI

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    DEFF Research Database (Denmark)

    Kaplan, Sigal; Prato, Carlo Giacomo

    2012-01-01

    Introduction: Recent years have witnessed a growing interest in improving bus safety operations worldwide. While in the United States buses are considered relatively safe, the number of bus accidents is far from being negligible, triggering the introduction of the Motor-coach Enhanced Safety Act...... 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...... 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...

  13. Assessment and limitation of radioactivity transfers in the event of a postulated severe PWR accident

    International Nuclear Information System (INIS)

    Gauvain, J.

    1992-01-01

    This report constitutes the supporting material for a lecture on severe accidents which could occur on PWR type nuclear reactors. It is assumed for present purposes that the reader has at least a rudimentary acquaintance with the basics of general physics if not with the operating processes of these reactors. After defining what is meant by a ''severe accident'' on a reactor, the possible phenomenology of such an accident is qualitatively described: loss of coolant and loss of containment integrity. A certain number of elements are then given for the quantitative assessment of these phenomena involving possible radioactivity transfers within and outside the plant. In conclusion, available means are indicated for the limitation and control of these environmental transfers. (author). 5 refs, figs

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

    International Nuclear Information System (INIS)

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

    2000-01-01

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

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

    International Nuclear Information System (INIS)

    2013-01-01

    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)

  16. Application of nanofluids to mitigation of severe accident at the Fukushima Daiichi Nuclear Power Plant

    International Nuclear Information System (INIS)

    Suh, Kune Y.

    2011-01-01

    This paper presents the pivotal thermohydrodynamic and neutronic characteristics of nanofluids as an alternative coolant for boiling water reactors (BWRs) during the abnormal operation including, but not necessarily limited to, severe accident involving meltdown as well as potentially melt-through of the reactor core material. Results indicate that the benefit of utilizing nanoparticles in the BWR working fluid appears to be minimal during the nominal operation since the nanoparticles tend to carry over to the turbine and condenser lending themselves to erosion and fouling concerns. Good news, on the other hand, is that exploitation of nanofluids during the decay heat removal condition in case of an accident is promising indeed because of their high thermal conductivity and their neutron poisoning effect. Thermohydrodynamic and neutronic investigations are in progress to streamline the nanoparticles and to optimize their concentration during the abnormal operation beyond the conventional design basis extending to severe accident. (author)

  17. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2004-07-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

  18. Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

    International Nuclear Information System (INIS)

    Lee, Young Jin; Chung, Bub Dong

    2004-01-01

    MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

  19. Study on protective layer for severe accident conditions for EC6 reactor vault structure

    Energy Technology Data Exchange (ETDEWEB)

    Abrishami, H.H.; Ricciuti, R.; Khan, A.; Singh, R. [Candu Energy Inc., Mississauga, Ontario (Canada)

    2013-09-15

    The Enhanced CANDU 6 (EC6) is designed both for the prevention and mitigation of Design Basis Accidents (DBAs) as well as Beyond Design Basis Accidents (BDBAs). The foremost objective, in accordance with the safety goals specified in the CNSC Regulatory Document (RD-337), is to prevent the occurrence of any accident that could jeopardize nuclear safety, and, if an accident should occur, to limit the radiological releases resulting from the accident and minimize the impact on nearby communities. During a postulated severe core accident, Molten Core-Concrete Interaction (MCCI) may occur when molten core debris breaches the calandria vessel and contacts concrete surfaces, whereby the thermal and chemical properties of the melt contribute to the potential degradation of the concrete. The earliest phase of MCCI is characterized by very-high-temperature molten metal and oxide pouring from the calandria vessel and settling as a pool on the concrete surfaces of the vault floor. The molten material can result in spalling or fragmentation of the concrete near where the corium first contacts the concrete. As the corium settles on the concrete surface, the melt begins to react chemically with the concrete through the penetrating cracks and fragments produced on the initial contact, generating various gases including carbon monoxide and combustible hydrogen. In order to control and mitigate MCCI, a protective layer (refractory material) with suitable material properties and sufficient thickness was proposed to protect the reactor vault concrete floor. To further enhance vault floor protection and mitigate the conditions under severe accidents a special concrete composition in the upper layer of the vault floor concrete is to be provided in case the refractory material is breached. This special concrete should minimize the generation of various gases including combustible hydrogen and carbon monoxide during MCCI. As a part of research and development program an experimental

  20. Study on protective layer for severe accident conditions for EC6 reactor vault structure

    Energy Technology Data Exchange (ETDEWEB)

    Abrishami, H.H.; Ricciuti, R.; Khan, A.; Singh, R., E-mail: homayoun.abrishami@candu.com [Caandu Energy Inc., Mississauga, Ontario, (Canada)

    2013-07-01

    The Enhanced CANDU 6 (EC6) is designed both for the prevention and mitigation of Design Basis Accidents (DBAs) as well as Beyond Design Basis Accidents (BDBAs). The foremost objective, in accordance with the safety goals specified in the CNSC Regulatory Document (RD-337), is to prevent the occurrence of any accident that could jeopardize nuclear safety, and, if an accident should occur, to limit the radiological releases resulting from the accident and minimize the impact on nearby communities. During a postulated severe core accident, Molten Core-Concrete Interaction (MCCI) may occur when molten core debris breaches the calandria vessel and contacts concrete surfaces, whereby the thermal and chemical properties of the melt contribute to the potential degradation of the concrete. The earliest phase of MCCI is characterized by very-high-temperature molten metal and oxide pouring from the calandria vessel and settling as a pool on the concrete surfaces of the vault floor. The molten material can result in spalling or fragmentation of the concrete near where the corium first contacts the concrete. As the corium settles on the concrete surface, the melt begins to react chemically with the concrete through the penetrating cracks and fragments produced on the initial contact, generating various gases including carbon monoxide and combustible hydrogen. In order to control and mitigate MCCI, a protective layer (refractory material) with suitable material properties and sufficient thickness was proposed to protect the reactor vault concrete floor. To further enhance vault floor protection and mitigate the conditions under severe accidents a special concrete composition in the upper layer of the vault floor concrete is to be provided in case the refractory material is breached. This special concrete should minimize the generation of various gases including combustible hydrogen and carbon monoxide during MCCI. As a part of research and development program an experimental

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-03-15

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

  3. Teaching of severe accident of Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power

    International Nuclear Information System (INIS)

    Saito, Shinzo

    2011-01-01

    The Great East Japan Earthquake and accompanied tsunami brought about the severe accident at Fukushima Daiichi Nuclear Power Plants of Tokyo Electric Power Co., Inc. For 'No more Fukushima', twelve teaching of the accident was pointed out as follows: 1) natural disasters and external events shall be taken into consideration, 2) severe accident shall be included into safety regulation, 3) all possibility of hydrogen explosion shall be excluded, 4) diversity of safety important component and equipment shall be added with sufficient period of outage, 5) siting of multiple units at the same site shall be avoided at quake-prone country like Japan, 6) accident response environment for operators shall be improved, 7) accident convergence termination system shall be established so as to concentrate technical experience and knowledge, 8) off-site center shall be improved, 9) resident evacuation, consumption limit of food, radiation exposure and soil contamination limit shall be decided openly, 10) nuclear regulation and prevention of disaster shall be conducted by unitary organization to gain public trust, 11) fostering of safety culture among relevant enterprises shall be more encouraged and 12) nuclear industry shall develop reactor such as with no core meltdown or no evacuation and environmental contamination even if reactor core would be meltdown. (T. Tanaka)

  4. A simple assessment scheme for severe accident consequences using release parameters

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Kampanart, E-mail: kampanarts@tint.or.th [Thailand Institute of Nuclear Technology, 16 Vibhavadi-Rangsit Rd., Latyao, Chatuchak, 10900 (Thailand); Okamoto, Koji [The University of Tokyo, 7-3-1 Hongo, Bunkyo, Tokyo 113-8654 (Japan)

    2016-08-15

    Highlights: • Nuclear accident consequence index can assess overall consequences of an accident. • Correlations between the index and release parameters are developed. • Relation between the index and release amount follows power function. • The exponent of the power function is the key to the relation. - Abstract: Nuclear accident consequence index (NACI) which can assess the overall consequences of a severe accident on people and the environment is developed based on findings from previous studies. It consists of three indices: radiation effect index, relocation index and decontamination index. Though the NACI can cover large range of consequences, its assessment requires extensive resources. The authors then attempt to simplify the assessment, by investigating the relations between the release parameters and the NACI, in order to use the release parameters for severe accident consequence assessment instead of the NACI. NACI and its components increase significantly when the release amount is increased, while the influences of the release period and the release starting time on the NACI are nearly negligible. Relations between the release amount and the NACI and its components follow simple power functions (y = ax{sup b}). The exponent of the power functions seems to be the key to the relations. The exponent of the relation between the release amount and the NACI was around 0.8–1.0 when the release amount is smaller than 100 TBq, and it increased to around 1.3–1.4 when the release amount is equal to or larger than 100 TBq.

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

    International Nuclear Information System (INIS)

    Rashad, S.M.

    2013-01-01

    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

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

    International Nuclear Information System (INIS)

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

    2011-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  8. 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.

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

    International Nuclear Information System (INIS)

    1998-03-01

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

  10. Rupther: a simulation approach applied to a PWR vessel failure during a severe accident

    International Nuclear Information System (INIS)

    Mongabure, Ph.; Nicolas, L.; Devos, J.

    2000-01-01

    The Rupther program (Rupture Under Thermal Conditions) is a part of the international researches on severe accidents in the PWR type reactors. The aim of the program is the definition of failure simulation validated by experimental data on vessel steel 16MND5 mechanical properties. The paper presents the program and the first results. (A.L.B.)

  11. Decision model support of severity of injury traffic accident victims care by SAMU 192

    Directory of Open Access Journals (Sweden)

    Rackynelly Alves Sarmento Soares

    2013-01-01

    Full Text Available Traffic accidents produce high morbidity and mortality in several countries, including Brazil. The initial care to victims of accidents, by a specialized team, has tools for evaluating the severity of trauma, which guide the priorities. This study aimed to develop a decision model applied to pre-hospital care, using the Abbreviated Injury Scale, to define the severity of the injury caused by the AT, as well to describe the features of accidents and their victims, occurred in Joao Pessoa, Paraiba. This is a descriptive epidemiological investigation, sectional, which analyzed all victims of traffic accidents attended by the SAMU 192, João Pessoa-PB, in January, April and June 2010. Data were collected in the medical regulation sheets of SAMU 192. Most of victims were male (76%, aged between 20 and 39 years (60%. Most injuries were classified as AIS1 (62.5%. The model of decision support implemented was the decision tree that managed to correctly classify 95.98% of the severity of injuries. By this model, it was possible to extract 29 rules of gravity classification of injury, which may be used for decision-making teams of the SAMU 192.

  12. Analysis of the behaviour of the Kozloduy NPP Unit 3 under severe accident conditions

    International Nuclear Information System (INIS)

    Velev, V.; Saraeva, V.

    2004-01-01

    The objective of the analysis is to study the behaviour of the Kozloduy NPP Unit 3 under severe accident conditions. The analysis is performed using computer code MELCOR 1.8.4. This report includes a brief description of Unit 3 active core as well as description and comparison of the key events

  13. Investigation on accident management measures for VVER-1000 reactors

    International Nuclear Information System (INIS)

    Tusheva, P.; Schaefer, F.; Rohde, U.; Reinke, N.

    2009-01-01

    A consequence of a total loss of AC power supply (station blackout) leading to unavailability of major active safety systems which could not perform their safety functions is that the safety criteria ensuring a secure operation of the nuclear power plant would be violated and a consequent core heat-up with possible core degradation would occur. Currently, a study which examines the thermal-hydraulic behaviour of the plant during the early phase of the scenario is being performed. This paper focuses on the possibilities for delay or mitigation of the accident sequence to progress into a severe one by applying Accident Management Measures (AMM). The strategy 'Primary circuit depressurization' as a basic strategy, which is realized in the management of severe accidents is being investigated. By reducing the load over the vessel under severe accident conditions, prerequisites for maintaining the integrity of the primary circuit are being created. The time-margins for operators' intervention as key issues are being also assessed. The task is accomplished by applying the GRS thermal-hydraulic system code ATHLET. In addition, a comparative analysis of the accident progression for a station blackout event for both a reference German PWR and a reference VVER-1000, taking into account the plant specifics, is being performed. (authors)

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

    International Nuclear Information System (INIS)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T.; Drouin, M.

    1995-01-01

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

  15. 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.

  16. 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.

  17. Impact of spatial kinetics in severe accident analysis for a large HWR

    International Nuclear Information System (INIS)

    Morris, E.E.

    1994-01-01

    The impact on spatial kinetics on the analysis of severe accidents initiated by the unprotected withdrawal of one or more control rods is investigated for a large heavy water reactor. Large inter- and intra-assembly power shifts are observed, and the importance of detailed geometrical modeling of fuel assemblies is demonstrated. Neglect of space-time effects is shown to lead to erroneous estimates of safety margins, and of accident consequences in the event safety margins are exceeded. The results and conclusions are typical of what would be expected for any large, loosely coupled core

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

    International Nuclear Information System (INIS)

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

    1988-01-01

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

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

    International Nuclear Information System (INIS)

    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

  20. Cost analysis and financial risk profile for severe reactor accidents at Waterford-3

    International Nuclear Information System (INIS)

    Cutbush, J.D.; Abbott, E.C.; Carpenter, W.L. Jr.

    1992-01-01

    To support Louisiana Power and Light Company (LP and L) in determining an appropriate level of nuclear property insurance for Waterford Steam Electric Station, Unit 3 (Waterford-3), ABZ, Incorporated, performed a series of cost analyses and developed a financial risk profile. This five-month study, conducted in 1991, identified the potential Waterford-3 severe reactor accidents and described each from a cleanup perspective, estimated the cost and schedule to cleanup from each accident, developed a probability distribution of associated financial exposure, and developed a profile of financial risk as a function of insurance coverage

  1. Occurrence and countermeasures of urban power grid accident

    Science.gov (United States)

    Wei, Wang; Tao, Zhang

    2018-03-01

    With the advance of technology, the development of network communication and the extensive use of power grids, people can get to know power grid accidents around the world through the network timely. Power grid accidents occur frequently. Large-scale power system blackout and casualty accidents caused by electric shock are also fairly commonplace. All of those accidents have seriously endangered the property and personal safety of the country and people, and the development of society and economy is severely affected by power grid accidents. Through the researches on several typical cases of power grid accidents at home and abroad in recent years and taking these accident cases as the research object, this paper will analyze the three major factors that cause power grid accidents at present. At the same time, combining with various factors and impacts caused by power grid accidents, the paper will put forward corresponding solutions and suggestions to prevent the occurrence of the accident and lower the impact of the accident.

  2. Analysis of effects of calandria tube uncovery under severe accident conditions in CANDU reactors

    International Nuclear Information System (INIS)

    Rogers, J.T.; Currie, T.C.; Atkinson, J.C.; Dick, R.

    1983-01-01

    A study is being undertaken for the Atomic Energy Control Board to assess the thermal and hydraulic behaviour of CANDU reactor cores under accident conditions more severe than those normally considered in the licensing process. In this paper, we consider the effects on a coolant channel of the uncovery of a calandria tube by moderator boil-off following a LOCA in a Bruce reactor unit in which emergency cooling is ineffective and the moderator heat sink is impaired by the failure of the moderator cooling system. Calandria tube uncovery and its immediate consequences, as described here, constitute only one part of the entire accident sequence. Other aspects of this sequence as well as results of the analysis of the other accident sequences studied will be described in the final report on the project and in later papers

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

    Science.gov (United States)

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

    2018-02-01

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

  4. The circumstances of severe accident measure implementation and 'the residual risk'

    International Nuclear Information System (INIS)

    Hirano, Mitsumasa

    2011-01-01

    Time-series sequence and direct and root causes of Fukushima Daiichi accident were up to validation of Hatamura's investigation committee on the accident but it would be clear that measure against tsunamis was not good enough. Based on this unprecedented accident, revision of safety design review guide and regulatory requirements of severe accident (SA) measure were under consideration while SA measure had been implemented as public self-safety management by administrative guidance. History of SA measure preparation including the introduction of 'the residual risk' for expansion and upgrade of SA measure in new review guide of seismic design of nuclear power reactor facilities was looked back to learn lessons for better safety operation of nuclear facilities. Nuclear operators established accident management (AM) incorporating appropriate SA measure extracted from probabilistic safety assessment (PSA) in 2002, which had been expanded and reinforced by periodic safety review (PSR). At the revision of regulation in 2003, PSA became requirement of operational safety program but not mandatory as before and lost the chance of regulatory review at the PSR. Extent of SA measure had not been expanded based on latest knowledge of SA research and PSA technology. Evaluation of 'the residual risk' obtained by seismic PSA could not be reported at seismic back check so far because seismic evaluation against ground motion was obliged to be preferred. Safety regulation system based on safety culture of both nuclear operators and regulators should be established for implementation of advanced AM for a certainty. (T. Tanaka)

  5. 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.

  6. Study of the iodine kinetics in thermal conditions of a RCS in nuclear severe accident

    International Nuclear Information System (INIS)

    Grégoire, A.-C.; Délicat, Y.; Tornabene, C.

    2017-01-01

    Highlights: • High temperature iodine reactivity in gas phase. • Kinetics limitations were evidenced in severe accident gas atmosphere conditions. • Iodine kinetic scheme involving steam/di-hydrogen/di-oxygen. - Abstract: During the PHEBUS-FP integral severe accidents simulation tests, gaseous iodine was detected in earlier stages of the simulated accident, coming from the experimental circuit modelling a reactor coolant system. One possible explanation is the existence of some kinetic limitations which promote the persistence of gaseous iodine at low temperature. This paper sums up some analytical and modelling works performed to check this assumption. Results show that the chemical speciation of iodine cannot be calculated by assuming chemical equilibrium, kinetics have to be considered, in particular for oxidising atmosphere with an excess of steam. A kinetic model for gaseous iodine is proposed and qualified by comparison with experimental works. Such modelling should be considered to calculate the transport of iodine through the reactor coolant system for a severe accident because it can significantly impact iodine source term evaluations.

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

    International Nuclear Information System (INIS)

    Lee Min; Ko, Y.-C.

    2008-01-01

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

  8. 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

  9. 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

  10. Conclusions of the specialist meeting on operator AIDS for severe accident management and training (SAMOA)

    International Nuclear Information System (INIS)

    1994-01-01

    The scope of the Specialist Meeting was limited to operator aids for accident management which were in operation or could be soon. Moreover, 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. The presentations showed that the design and implementation of operator aids were closely related to the organisation adopted by the user, whether it was a utility or a governmental agency. The most common organisation is to share the management of severe accidents among two groups of people: the operating team in the Control Room (CR) and a team of specialists in a Technical Support Centre (TSC). The CR is in charge of the operation of the plant in all conditions using a set of procedures and guidelines, while the experts in the TSC are able to produce in-depth analyses of the plant state and its evolution. The responsibility is shared between the CR and the TSC during accident progression. The TSC acts as a support for the CR for reactor operation and takes charge of the predictions of radioactive releases (source term, accident progression, release and dispersion of radioactive substances, as well as the interaction with public authorities). But this type of organisation is not general and the differences can induce different approaches in the design of operator aids. The first session was dedicated to operator aids for control rooms, the second session to operator aids for technical support centres

  11. Large Break LOCA Accident Management Strategies for Accidents With Large Containment Leaks

    International Nuclear Information System (INIS)

    Sdouz, Gert

    2006-01-01

    The goal of this work is the investigation of the influence of different accident management strategies on the thermal-hydraulics in the containment during a Large Break Loss of Coolant Accident with a large containment leak from the beginning of the accident. The increasing relevance of terrorism suggests a closer look at this kind of severe accidents. Normally the course of severe accidents and their associated phenomena are investigated with the assumption of an intact containment from the beginning of the accident. This intact containment has the ability to retain a large part of the radioactive inventory. In these cases there is only a release via a very small leakage due to the un-tightness of the containment up to cavity bottom melt through. This paper represents the last part of a comprehensive study on the influence of accident management strategies on the source term of VVER-1000 reactors. Basically two different accident sequences were investigated: the 'Station Blackout'- sequence and the 'Large Break LOCA'. In a first step the source term calculations were performed assuming an intact containment from the beginning of the accident and no accident management action. In a further step the influence of different accident management strategies was studied. The last part of the project was a repetition of the calculations with the assumption of a damaged containment from the beginning of the accident. This paper concentrates on the last step in the case of a Large Break LOCA. To be able to compare the results with calculations performed years ago the calculations were performed using the Source Term Code Package (STCP), hydrogen explosions are not considered. In this study four different scenarios have been investigated. The main parameter was the switch on time of the spray systems. One of the results is the influence of different accident management strategies on the source term. In the comparison with the sequence with intact containment it was

  12. Workshop on iodine aspects of severe accident management. Summary and conclusions

    International Nuclear Information System (INIS)

    2000-03-01

    Following a recommendation of the OECD Workshop on the Chemistry of Iodine in Reactor Safety held in Wuerenlingen (Switzerland) in June 1996 [Summary and Conclusions of the Workshop, Report NEA/CSNI/R(96)7], the CSNI decided to sponsor a Workshop on Iodine Aspects of Severe Accident Management, and their planned or effective implementation. The starting point for this conclusion was the realization that the consolidation of the accumulated iodine chemistry knowledge into accident management guidelines and procedures remained, to a large extent, to be done. The purpose of the meeting was therefore to help build a bridge between iodine research and the application of its results in nuclear power plants, with particular emphasis on severe accident management. Specifically, the Workshop was expected to answer the following questions: - what is the role of iodine in severe accident management? - what are the needs of the utilities? - how can research fulfill these needs? The Workshop was organized in Vantaa (Helsinki), Finland, from 18 to 20 May 1999, in collaboration with Fortum Engineering Ltd. It was attended by forty-six specialists representing fifteen Member countries and the European Commission. Twenty-eight papers were presented. These included four utility papers, representing the views of Electricite de France (EDF), Teollisuuden Voima Oy and Fortum Engineering Ltd (Finland), the Nuclear Energy Institute (USA), and Japanese utilities. The papers were presented in five sessions: - iodine speciation; - organic compound control; - iodine control; - modeling; - iodine management; A sixth session was devoted to a general discussion on iodine management under severe accident conditions. This report summarizes the content of the papers and the conclusions of the workshop

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-05-15

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

  14. Assessment of the potential for HPME during a station blackout in the Surry and Zion PWRS

    International Nuclear Information System (INIS)

    Knudson, D.L.; Bayless, P.D.; Dobbe, C.A.; Odar, F.

    1994-01-01

    The integrity of a PWR (pressurized water reactor) containment structure could be challenged by direct heating associated with a HPME (high pressure melt ejection) of core materials following reactor vessel lower head breach during certain severe accidents. Structural failure resulting from direct containment heating is a contributor to the risk of operating a PWR. Intentional RCS (reactor coolant system) depressurization, where operators latch pressurizer relief valves open, has been proposed as an accident management strategy to reduce those risks by mitigating the severity of the HPME. However, decay heat levels, valve capacities, and other plant-specific characteristics determine whether the required operator action will be effective. Without operator action, natural circulation flows could heat ex-vessel RCS pressure boundaries (surge line and hot leg piping, steam generator tubes, etc.) to the point of failure before failure of the lower head providing an unintentional mechanism for depressurization and HPME mitigation. This paper summarizes an assessment of RCS depressurization with respect to the potential for HPME during a station blackout in the Surry and Zion PWRs. The assessment included a detailed transient analysis using the SCDAP/RELAP5/MOD3 computer code and an evaluation of RCS depressurization-related probabilities primarily based on the code results

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

    International Nuclear Information System (INIS)

    2015-11-01

    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

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

    International Nuclear Information System (INIS)

    2014-01-01

    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

  17. Common Risk Target for severe accidents of nuclear power plants based on IAEA INES scale

    International Nuclear Information System (INIS)

    Vitázková, Jiřina; Cazzoli, Errico

    2013-01-01

    The IAEA has repeatedly recommended that the nuclear community should arrive at a common understanding and definition of safety goals for severe accidents in nuclear power plants. The recommendation has only found partial answers, despite the numerous working groups and forums devoted to this effort. The most widely accepted definition of goals is based on the concept of Large (Early) Release Frequencies (L(E)RF) and its derivatives, a surrogate concept derived from results of Probabilistic Safety Assessments (PSAs) which was first introduced in the USA almost twenty years ago and much later accepted by the USNRC for risk informed decision making, but not for safety demonstrations. Other types of Safety Goals have been adopted by some nuclear authorities, but the main drawback of all current definitions is that they may apply only to LWRs. The lack of unifying safety/risk parameter throughout of PSAs worldwide is the basis of the present work, and an attempt is made to arrive at the definition of a Risk Target for severe accidents in NPPs, consistent with the IAEA definitions having a technical basis, which can be adopted without modifications for Generation IV power plants. The proposal of Common Risk Target in this work represents an attempt to define a Common Risk Target based on technical reasoning, reflecting IAEA definitions as well as harmonization requirements raised by the whole European Community in various OECD, ASAMPSA2 and SARNET (Guentay et al., 2006) conclusions and Council Directive of The European Union (Community Framework, 2009) as well as lastly performed stress tests of nuclear power plants throughout the Europe (Peer Review Report, 2012). The basic concept of CRT was first introduced and developed within the European project ASAMPSA2 by the authors of this article and was accepted by majority of world PSA experts participating in final evaluation and survey of the project (Guentay, 2011). In the proposed Risk Target concept an innovative

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

    International Nuclear Information System (INIS)

    Jae, M.; Moon, J.H.

    2002-01-01

    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

  19. Assessment of severe accident prevention and mitigation features: PWR, large dry containment design

    International Nuclear Information System (INIS)

    Perkins, K.R.; Hsu, C.J.; Lehner, J.R.; Luckas, W.J.; Cho, N.; Fitzpatrick, R.G.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

    Plant features and operator actions which have been found to be important in either preventing or mitigating severe accidents in PWRs with large dry containments have been identified. These features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Zion plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the large dry containment to severe accident containment loads were also identified. In addition, those features of a PWR with a large dry containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. The report is issued to provide focus to the analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic tributes for assessing those features and actions found to be helpful in reducing the overall risk for Zion and other PWRs with large dry containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  20. Assessment of severe accident prevention and mitigation features: PWR, ice-condenser containment design

    International Nuclear Information System (INIS)

    Hsu, C.J.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Cho, N.; Lehner, J.R.; Pratt, W.T.; Eltawila, F.; Maly, J.A.

    1988-07-01

    Plant features and operator actions which have been found to be important in either preventing and mitigating severe accidents in PWRs with ice-condenser containments have been identified. Thus features and actions were developed from insights derived from reviews of risk assessments performed specifically for the Sequoyah plant and from assessments of other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the ice-condenser containment to sever accident containment loads were also identified. In addition, those features of a PWR with an ice-condenser containment, which are important for preventing core damage and are available for mitigating fission-product release to the environment were identified. This report is issued to provide focus to an analyst examining an individual plant. The report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Sequoyah and other PWRs with ice-condenser containments. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance. 14 tabs

  1. Assessment of severe accident prevention and mitigation features: BWR, Mark II containment design

    International Nuclear Information System (INIS)

    Lehner, J.R.; Hsu, C.J.; Eltawila, F.; Perkins, K.R.; Luckas, W.J.; Fitzpatrick, R.G.; Pratt, W.T.

    1988-07-01

    Plant features and operator actions, which have been found to be important in either preventing or mitigating severe accidents in BWRs with Mark II containments (BWR Mark II's) have been identified. These features and actions were developed from insights derived from reviews of in-depth risk assessments performed specifically for the Limerick and Shoreham plants and from other relevant studies. Accident sequences that dominate the core-damage frequency and those accident sequences that are of potentially high consequence were identified. Vulnerabilities of the BWR Mark II to severe-accident containment loads were also noted. In addition, those features of a BWR Mark II, which are important for preventing core damage and are available for mitigating fission-product release to the environment were also identified. This report is issued to provide focus to an analyst examining an individual plant. This report calls attention to plant features and operator actions and provides a list of deterministic attributes for assessing those features and actions found to be helpful in reducing the overall risk for Mark II plants. Thus, the guidance is offered as a resource in examining the subject plant to determine if the same, or similar, plant features and operator actions will be of value in reducing overall plant risk. This report is intended to serve solely as guidance

  2. 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.

  3. [Severe perforating eye injury caused by an air bag in a traffic skid accident].

    Science.gov (United States)

    Biechl-Lautenbach, K S; Gloor, B; Walz, F

    1996-03-01

    After a penetrating injury of the eye following a car accident the driver demanded accurate investigations by the Institute for Legal Medicine of Zurich concerning the security of the equipment, especially the airbag. This led to an astonishing explication of the mechanism of the injury, not without consequences for safety measures for drivers and front seat passengers in air bag equipped cars. Due to a relatively harmless accident, the driver suffered from a severe penetrating injury of the right eye after the airbag deployed. The front seat passenger, having the seat belts fastened, was not injured. The accident was investigated by the Institute of Legal Medicine of Zurich. The analysis of the accident showed that the airbag had deployed properly. The cover of the airbag showed no defects of substance. With the precise examination of the interior of the car a broken tobacco pipe came to light. In this case not the airbag itself but a tobacco pipe held in the hand by the driver during the airbag ignition caused a severe injury of the eye. This case report illustrates the hazard of having any rigid object between the occupant and the deploying air bag. In conclusion, the drivers and front seat passengers of an airbag equipped car can only profit from the considerable security gain, if they know about these risks and adapt their behaviours to the new surroundings, but also the car manufactures have to instruct the customers properly.

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

    International Nuclear Information System (INIS)

    1997-01-01

    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

  5. 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.

  6. Station blackout core damage frequency in an advanced nuclear reactor

    International Nuclear Information System (INIS)

    Carvalho, Luiz Sergio de

    2004-01-01

    Even though nuclear reactors are provided with protection systems so that they can be automatically shut down in the event of a station blackout, the consequences of this event can be severe. This is because many safety systems that are needed for removing residual heat from the core and for maintaining containment integrity, in the majority of the nuclear power plants, are AC dependent. In order to minimize core damage frequency, advanced reactor concepts are being developed with safety systems that use natural forces. This work shows an improvement in the safety of a small nuclear power reactor provided by a passive core residual heat removal system. Station blackout core melt frequencies, with and without this system, are both calculated. The results are also compared with available data in the literature. (author)

  7. Enhanced safety features of CHASHMA NPP UNIT-2 to encounter selected severe accidents, various challenges involved to prove the adequacy of severe accidents prevention/mitigation measures and to write management guidelines with one possible solution to these challenges

    International Nuclear Information System (INIS)

    Iqbal, Z.; Minhaj, A.

    2007-01-01

    This paper describes enhanced safety features of Chashma Nuclear Power Plant Unit-2 (C-2), a 325 MWe PWR to encounter selected severe accidents and discusses various challenges involved to prove the adequacy of severe accidents encountering measures and to write severe accident management guidelines (SAMGs) in compliance with the recently introduced national regulations based on the new IAEA nuclear safety standards. C-2 is being built by China National Nuclear Corporation (CNNC) for Pakistan Atomic Energy Commission (PAEC). Its twin, Unit-1 (C-1) also a 325 MWe PWR, was commissioned in 2000. Nuclear power safety with reference to severe accidents should be treated as a global issue and therefore the developed countries should include the people of developing countries in nuclear power industry's various severe accidents based research and development programs. The implementation of this idea may also deliver few other useful and mutually beneficial byproducts. (author)

  8. Introduction to Large-sized Test Facility for validating Containment Integrity under Severe Accidents

    International Nuclear Information System (INIS)

    Na, Young Su; Hong, Seongwan; Hong, Seongho; Min, Beongtae

    2014-01-01

    An overall assessment of containment integrity can be conducted properly by examining the hydrogen behavior in the containment building. Under severe accidents, an amount of hydrogen gases can be generated by metal oxidation and corium-concrete interaction. Hydrogen behavior in the containment building strongly depends on complicated thermal hydraulic conditions with mixed gases and steam. The performance of a PAR can be directly affected by the thermal hydraulic conditions, steam contents, gas mixture behavior and aerosol characteristics, as well as the operation of other engineering safety systems such as a spray. The models in computer codes for a severe accident assessment can be validated based on the experiment results in a large-sized test facility. The Korea Atomic Energy Research Institute (KAERI) is now preparing a large-sized test facility to examine in detail the safety issues related with hydrogen including the performance of safety devices such as a PAR in various severe accident situations. This paper introduces the KAERI test facility for validating the containment integrity under severe accidents. To validate the containment integrity, a large-sized test facility is necessary for simulating complicated phenomena induced by an amount of steam and gases, especially hydrogen released into the containment building under severe accidents. A pressure vessel 9.5 m in height and 3.4 m in diameter was designed at the KAERI test facility for the validating containment integrity, which was based on the THAI test facility with the experimental safety and the reliable measurement systems certified for a long time. This large-sized pressure vessel operated in steam and iodine as a corrosive agent was made by stainless steel 316L because of corrosion resistance for a long operating time, and a vessel was installed in at KAERI in March 2014. In the future, the control systems for temperature and pressure in a vessel will be constructed, and the measurement system

  9. Utilities respond to nuclear station blackout rule

    International Nuclear Information System (INIS)

    Rubin, A.M.; Beasley, B.; Tenera, L.P.

    1990-01-01

    The authors discuss how nuclear plants in the United States have taken actions to respond to the NRC Station Blackout Rule, 10CFR50.63. The rule requires that each light water cooled nuclear power plant licensed to operate must be able to withstand for a specified duration and recover from a station blackout. Station blackout is defined as the complete loss of a-c power to the essential and non-essential switch-gear buses in a nuclear power plant. A station blackout results from the loss of all off-site power as well as the on-site emergency a-c power system. There are two basic approaches to meeting the station blackout rule. One is to cope with a station blackout independent of a-c power. Coping, as it is called, means the ability of a plant to achieve and maintain a safe shutdown condition. The second approach is to provide an alternate a-c power source (AAC)

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

    International Nuclear Information System (INIS)

    Poumerouly, S.

    2010-01-01

    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

  11. 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. Copyright © 2015 Elsevier Ltd. All rights reserved.

  12. Independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool

    International Nuclear Information System (INIS)

    Madni, I.K.; Eltawila, F.

    1994-01-01

    MELCOR is a fully integrated computer code that models all phases of the progression of severe accidents in light water reactor nuclear power plants, and is being developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL). Brookhaven National Laboratory (BNL) has a program with the NRC called ''MELCOR Verification, Benchmarking, and Applications,'' whose aim is to provide independent assessment of MELCOR as a severe accident thermal-hydraulic/source term analysis tool. The scope of this program is to perform quality control verification on all released versions of MELCOR, to benchmark MELCOR against more mechanistic codes and experimental data from severe fuel damage tests, and to evaluate the ability of MELCOR to simulate long-term severe accident transients in commercial LWRs, by applying the code to model both BWRs and PWRs. Under this program, BNL provided input to the NRC-sponsored MELCOR Peer Review, and is currently contributing to the MELCOR Cooperative Assessment Program (MCAP). This paper presents a summary of MELCOR assessment efforts at BNL and their contribution to NRC goals with respect to MELCOR

  13. Phenomenology and course of severe accidents in PWR-plants training by teaching and demonstration

    International Nuclear Information System (INIS)

    Sonnenkalb, M.; Rohde, J.

    1999-01-01

    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 (BMW) and for Education, Science, Research and Technology (BMBF). In the paper the intention and the subject of this training course are discussed and selected parts of the training course are 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 training course will be made available for members of crisis teams, instructors of commercial training centres and researchers of different institutions too. (author)

  14. Comparative assessment of severe accident risks in the coal, oil and natural gas chains

    International Nuclear Information System (INIS)

    Burgherr, Peter; Eckle, Petrissa; Hirschberg, Stefan

    2012-01-01

    This study compared severe accident risks of fossil energy chains (coal, oil and natural gas), based on the historical experience contained in the comprehensive database ENSAD. Considered risk indicators focused on human health impacts, i.e., fatality rates and maximum consequences were calculated for a broad range of country groups. Generally, expected fatality rates were lowest for natural gas, intermediate for oil and highest for coal. Concerning maximum consequences of a single accident, natural gas also performed best, followed by coal, whereas accidents in the oil chain can claim significantly more fatalities. In general, OECD and EU 27 ranked top, while non-OECD countries and China in the case of coal were worst. The consideration of numerous additional country groups enabled a more detailed differentiation within the main bounding groups. Furthermore, differences among country groups are distinctly decreasing from coal to oil and natural gas, both for fatality rates and maximum consequences. The use of import adjusted-fatality rates indicates that fatality risks in supply countries are an essential aspect to understand how specific risk reduction strategies may affect other components of energy security, and thus tradeoffs and compromises are necessary. Finally, the proposed fatality risk score for fossil chains (FRS F ) allows a comparison of the combined accident risk for the considered fossil energy chains across individual countries, which can be visualized using risk mapping.

  15. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-23

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

  17. The Severe Accident Mitigation Concept of Arena NPPs EPR{sup TM} Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, M. [AREVA NP, Erlangen (Germany)

    2012-03-15

    All AREVA NP Gen-3 reactors, the EPR{sup TM} and ATMEA1{sup TM} PWR, as well as KERENA BWR, give extended consideration to the prevention of severe accidents. Despite the so-achieved very low core melt frequencies, additional 'beyond-safety' measures are implemented in these designs. They are aimed at drastically reducing the environmental impact of a severe accident (SA), should it nevertheless occur, as well as at eliminating the need for emergency evacuations of the surrounding population and long-term restrictions with respect to the consumption of LOCAy grown food. The adopted safety concepts meet advanced regulatory requirements, incl. IAEA safety guide, Technical Guidelines (GPR/German experts), and SECY 93-087 (NRC). The chosen SA concept is illustrated at the example of the EPR{sup TM} plant which is currently under construction in Finland, France and China.

  18. Characterization of debris/concrete interactions for advanced research reactor and commercial BWR severe accidents

    International Nuclear Information System (INIS)

    Hyman, C.R.; Taleyarkhan, R.P.; Greene, S.R.

    1991-01-01

    The core concrete interaction (CCI) is an important phase of any severe accident where the reactor vessel has failed and core debris is relocated onto the containment basemat. In recent calculations performed at the Oak Ridge National Laboratory (ORNL), CCI has been studied for severe accidents occurring in a commercial Boiling Water Reactor (BWR) and in a high-power density Department of Energy (DOE) research reactor that is currently in the conceptual design stage. Because of differences in the debris decay heating level, core debris composition and inventory, and containment design, the characteristics of the resulting CCI and containment response are different for the two reactor types. Furthermore, proper selection of the basemat concrete type and the provision of an overlying water pool are found to be significant CCI mitigating factors for the research reactor and thus constitute important design considerations for any future reactor type. 10 refs., 4 figs., 1 tab

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

    International Nuclear Information System (INIS)

    Heams, T.J.; Williams, D.A.; Johns, N.A.; Mason, A.; Bixler, N.E.; Grimley, A.J.; Wheatley, C.J.; Dickson, L.W.; Osborn-Lee, I.; Domagala, P.; Zawadzki, S.; Rest, J.; Alexander, C.A.; Lee, R.Y.

    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

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

    International Nuclear Information System (INIS)

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

    2002-01-01

    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

  1. Pressure drop-flow rate curves for single-phase steam in Combustion Engineering type steam generator U-tubes during severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fynan, Douglas A.; Ahn, Kwang-Il, E-mail: kiahn@kaeri.re.kr

    2016-12-15

    Highlights: • Pressure drop-flow rate curves for superheated steam in U-tubes were generated. • Forward flow of hot steam is favored in the longer and taller U-tubes. • Reverse flow of cold steam is favored in short U-tubes. • Steam generator U-tube bundle geometry and tube diameter are important. • Need for correlation development for natural convention heat transfer coefficient. - Abstract: Characteristic pressure drop-flow rate curves are generated for all row numbers of the OPR1000 steam generators (SGs), representative of Combustion Engineering (CE) type SGs featuring square bend U-tubes. The pressure drop-flow rate curves are applicable to severe accident natural circulations of single-phase superheated steam during high pressure station blackout sequences with failed auxiliary feedwater and dry secondary side which are closely related to the thermally induced steam generator tube rupture event. The pressure drop-flow rate curves which determine the recirculation rate through the SG tubes are dependent on the tube bundle geometry and hydraulic diameter of the tubes. The larger CE type SGs have greater variation of tube length and height as a function of row number with forward flow of steam favored in the longer and taller high row number tubes and reverse flow favored in the short low row number tubes. Friction loss, natural convection heat transfer coefficients, and temperature differentials from the primary to secondary side are dominant parameters affecting the recirculation rate. The need for correlation development for natural convection heat transfer coefficients for external flow over tube bundles currently not modeled in system codes is discussed.

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

    International Nuclear Information System (INIS)

    Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

    1992-01-01

    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

  3. Synthesis of VERCORS and Phebus data in severe accident codes and applications.

    Energy Technology Data Exchange (ETDEWEB)

    Gauntt, Randall O.

    2010-04-01

    The Phebus and VERCORS data have played an important role in contemporary understanding and modeling of fission product release and transport from damaged LWR fuel. The data from these test programs have allowed improvement of MELCOR modeling of release and transport processes for both low enrichment uranium fuel as well as high burnup and MOX fuels. The following paper describes the derivation, testing and incorporation of improved radionuclide release models into the MELCOR severe accident code.

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

    OpenAIRE

    Rameez Rameezdeen; Abbas Elmualim

    2017-01-01

    The impact of heat stress on human health has been extensively studied. Similarly, researchers have investigated the impact of heat stress on workers? health and safety. However, very little work has been done on the impact of heat stress on occupational accidents and their severity, particularly in South Australian construction. Construction workers are at high risk of injury due to heat stress as they often work outdoors, undertake hard manual work, and are often project based and sub-contr...

  5. Probabilistic approach in treatment of deterministic analyses results of severe accidents

    International Nuclear Information System (INIS)

    Krajnc, B.; Mavko, B.

    1996-01-01

    Severe accidents sequences resulting in loss of the core geometric integrity have been found to have small probability of the occurrence. Because of their potential consequences to public health and safety, an evaluation of the core degradation progression and the resulting effects on the containment is necessary to determine the probability of a significant release of radioactive materials. This requires assessment of many interrelated phenomena including: steel and zircaloy oxidation, steam spikes, in-vessel debris cooling, potential vessel failure mechanisms, release of core material to the containment, containment pressurization from steam generation, or generation of non-condensable gases or hydrogen burn, and ultimately coolability of degraded core material. To asses the answer from the containment event trees in the sense of weather certain phenomenological event would happen or not the plant specific deterministic analyses should be performed. Due to the fact that there is a large uncertainty in the prediction of severe accidents phenomena in Level 2 analyses (containment event trees) the combination of probabilistic and deterministic approach should be used. In fact the result of the deterministic analyses of severe accidents are treated in probabilistic manner due to large uncertainty of results as a consequence of a lack of detailed knowledge. This paper discusses approach used in many IPEs, and which assures that the assigned probability for certain question in the event tree represent the probability that the event will or will not happen and that this probability also includes its uncertainty, which is mainly result of lack of knowledge. (author)

  6. WIND project tests and analysis on the integrity of small size pipe under severe accident condition

    International Nuclear Information System (INIS)

    Nakamura, Naohiko; Hashimoto, Kazuichiro; Maruyama, Yu; Igarashi, Minoru; Hidaka, Akihide; Sugimoto, Jun

    1996-01-01

    In a severe accident of a light water reactor(LWR), fission products (FPs) released from fuel rods will be transported to the primary cooling system piping as aerosol and some of them will be deposited on the inner surface of piping. In such conditions the primary cooling system piping might be subjected to both of elevated temperature load due to decay heat of FPs and pressure load, and as a consequence the integrity of piping might be threatened. The WIND (Wide Range Piping Integrity Demonstration) Project is being performed at Japan Atomic Energy Research Institute (JAERI) to investigate the FP aerosol behavior in reactor piping and the integrity of reactor piping under severe accident condition (K. Hashimoto et al., 1994, K. Hashimoto et al., 1995). In order to meet these two objectives, the Project comprises two test series: an aerosol behavior test series and a piping integrity test series. In the piping integrity test a straight stainless steel pipe is used to simulate a partial fraction of reactor piping under severe accident conditions. In parallel with conducting the tests, test analyses are performed with ABAQUS code (Hibbitt, Karlsson and Sorensen Inc. 1989) using the test conditions to investigate the behavior of straight pipe against thermal and pressure loads. This paper describes the comparison of the scoping piping integrity test results and the analysis results with ABAQUS

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.

    2007-04-01

    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

  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. Sustainable integration of EU research in severe accident phenomenology and management

    International Nuclear Information System (INIS)

    Van Dorsselaere, Jean-Pierre; Albiol, Thierry; Chaumont, Bernard; Haste, Tim; Journeau, Christophe; Meyer, Leonhard; Sehgal, Bal Raj; Schwinges, Bernd; Beraha, David; Annunziato, Alessandro; Zeyen, Roland

    2011-01-01

    Highlights: → The SARNET network gathers most worldwide actors involved in severe accident research. → It defines common research programmes for resolving the most important pending safety issues. → It optimises the use of the available European resources and constitutes sustainable research groups. → It disseminates the knowledge on severe accidents through education courses. → Knowledge produced is capitalized through physical models in the ASTEC simulation code. - Abstract: 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, between 2004 and 2008, 51 organizations 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 and methodologies for safety assessment. The Joint Programme of Activities consisted in: -Implementing an advanced communication tool for accessing all project information, fostering exchange of information, and managing documents; - Harmonizing and re-orienting the research programmes, and defining new ones; -Analyzing the experimental results provided by research programmes in order to elaborate a common understanding of relevant phenomena; -Developing the ASTEC code (integral computer code used to predict the NPP behaviour during a postulated SA) by capitalizing in terms of physical models the knowledge produced within SARNET; - Developing scientific databases, in which the results of research experimental programmes are stored in a common

  12. 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.

  13. 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.

  14. Evaluation of Coolant Injection Procedure in the Severe Accident Management Strategy of APR1400

    International Nuclear Information System (INIS)

    Cho, Yongjin; Lim, Kukhee; Song, Sungchu; Lee, Sukho; Hwang, Taesuk

    2013-01-01

    A coolant injection strategy in the severe accident management guideline (SAMG) of APR1400 relates to immediate coolant injection into RCS (Reactor Coolant System) or injection following the recovery of secondary coolant inventory. This strategy could play important role in accident mitigation and radiological consequences. In this study, appropriateness of the strategy was evaluated using MELCOR1.8.6 and several sensitivity studies of the key parameters were performed. Analysis for APR1400 using MELCOR 1.8.6 was performed to evaluate the effectiveness of accident management strategies and the following conclusions were identified. Sequential operation of secondary and RCS injection may not be the best strategy and the simultaneous injection of secondary and RCS injection could be more preferable. At least, the RCS injection should start before complete drainage of water in the safety injection tank using mobile pumps. In this study, the effectiveness of timing of operator action has been examined and the amount of injection flowrate needs to be studied in the future

  15. Phenomenological uncertainties in the suspended radionuclide concentrations in containment during severe LWR accidents

    International Nuclear Information System (INIS)

    Williams, D.C.; Murata, K.K.; Tills, J.L.

    1985-01-01

    CONTAIN, a code for integrated analysis of containment phenomenologies in complex LWR sever accident sequences, is being applied in a program for evaluating the uncertainties in USNRC-sponsored efforts to better define LWR accident source terms. The Surry TMLB sequence is studied in detail. Aerosol agglomeration uncertainties are found to contribute about an order of magnitude to the overall uncertainty in suspended radionuclides (almost entirely downward uncertainty; i.e., downward with respect to current base case estimates such as BMI-2104). Containment compartmentalization effects contribute substantial uncertainties in either direction, while effects due to complex multicomponent aerosol compositions contribute lesser, but still potentially significant uncertainties (mostly downward). Incomplete treatment of radionuclide decay chains can contribute factor-of-two upward uncertainties

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

    International Nuclear Information System (INIS)

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

    1987-02-01

    This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. 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

  17. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

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

    International Nuclear Information System (INIS)

    Feuerbach, R.

    2002-01-01

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

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

    International Nuclear Information System (INIS)

    Bui, V.A.

    1998-01-01

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

  20. Severe accident management; the approach in the USA. Applications of US methods in Europe. Other approaches in Europe

    International Nuclear Information System (INIS)

    Vayssier, G.

    1999-01-01

    In this lecture severe accident management, applications of US methods in Europe are presented. Author deals with historical perspective, US industry position to core melt accidents, method of Westinghouse owners group, method of Combustion Engineering owners group, method of Babcock and Wilcox Owners group, interaction with/inspection by the USNRC and with assessment of US SAMG methods

  1. Analyzing fault and severity in pedestrian-motor vehicle accidents in China.

    Science.gov (United States)

    Zhang, Guangnan; Yau, Kelvin K W; Zhang, Xun

    2014-12-01

    The number of pedestrian-motor vehicle accidents and pedestrian deaths in China surged in recent years. However, a large scale empirical research on pedestrian traffic crashes in China is lacking. In this study, we identify significant risk factors associated with fault and severity in pedestrian-motor vehicle accidents. Risk factors in several different dimensions, including pedestrian, driver, vehicle, road and environmental factors, are considered. We analyze 6967 pedestrian traffic accident reports for the period 2006-2010 in Guangdong Province, China. These data, obtained from the Guangdong Provincial Security Department, are extracted from the Traffic Management Sector-Specific Incident Case Data Report. Pedestrian traffic crashes have a unique inevitability and particular high risk, due to pedestrians' fragility, slow movement and lack of lighting equipment. The empirical analysis of the present study has the following policy implications. First, traffic crashes in which pedestrians are at fault are more likely to cause serious injuries or death, suggesting that relevant agencies should pay attention to measures that prevent pedestrians from violating traffic rules. Second, both the attention to elderly pedestrians, male and experienced drivers, the penalty to drunk driving, speeding, driving without a driver's license and other violation behaviors should be strengthened. Third, vehicle safety inspections and safety training sessions for truck drivers should be reinforced. Fourth, improving the road conditions and road lighting at night are important measures in reducing the probability of accident casualties. Fifth, specific road safety campaigns in rural areas, and education programs especially for young children and teens should be developed and promoted. Moreover, we reveal a country-specific factor, hukou, which has significant effect on the severity in pedestrian accidents due to the discrepancy in the level of social insurance/security, suggesting

  2. The development and demonstration of integrated models for the evaluation of severe accident management strategies - SAMEM

    International Nuclear Information System (INIS)

    Ang, M.L.; Peers, K.; Kersting, E.; Fassmann, W.; Tuomisto, H.; Lundstroem, P.; Helle, M.; Gustavsson, V.; Jacobsson, P.

    2001-01-01

    This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the

  3. Review on Core Designs for Prevention of Severe Accidents in SFRs

    International Nuclear Information System (INIS)

    Bae, Moohoon; Choi, Yong Won; Shin, Andong; Suh, Namduk

    2013-01-01

    Based on this characteristic of fast reactor core, potential impact of CDA (Core Disruptive Accident) caused by ATWS has been considered as an important safety issue, although it is extremely unlikely. In order to prevent and mitigate the severe accident, the fast reactor core has been designed with various safety features. In this paper, as a part of study to develop the domestic regulatory requirements and guidelines related to SFR core safety, international trends on safety features which have been considered in current SFR cores are reviewed. In order to develop the regulatory requirements and guidelines related to a SFR core design for prevention of CDA, the core safety features were reviewed. The safety features considered in current SFR cores have a function that prevents to progress into next step in accident sequences. The trends on current safety features are as follows: · 'passive shutdown systems' to prevent initiating events such as ULOF, UTOP, ULOHS, etc · 'core designs with low void effect' to prevent the large void reactivity insertion in initiating phase · 'specific provisions for core with conventional positive void effect' to prevent the core recriticality in transition phase Consequently, in regulatory review requirements and guidelines developed for SFR, contents for not only reduction of positive void effect but also features to ensure the safety of overall system should be reflected

  4. Analytical and experimental investigations of the passive heat transport in HTRs under severe accident conditions

    International Nuclear Information System (INIS)

    Rehm, W.; Barthels, H.; Jahn, W.; Cleveland, J.C.; Ishihara, M.

    1992-01-01

    Thermodynamic accident analyses have been performed with computer simulation models to investigate core heatup sequences, sensitivity analyses, power variations, anticipated transients without scram, and core displacement considerations for probabilistic safety analyses (PSA) of small gas-cooled high-temperature reactors (e.g. HTR-Module). In worst case considerations where not only a loss of the active heat removal system is assumed but also a loss of the vessel cooling system, the heat would be transported into the surrounding concrete structure. In such a case the concrete would act as a natural long-term intermediate heat storage dissipating the heat through the concrete surface. Large scale and reactor safety experiments have been performed to investigate passive heat transport mechanisms -- which can cooldown a HIR core during severe accident conditions -- for validation basis of computer simulation codes used for accident analyses. In general, the comparisons of experimental and analytical results with computer calculations of the heat transport codes are in good agreement

  5. Management of NPP severe accident by prevention of clad-softening

    International Nuclear Information System (INIS)

    Saxena, Anil Kumar; Limaye, Sanjay Prabhakar; Bera, Subrata; Deo, Anuj Kumar

    2015-01-01

    Specified Emergency Core Cooling System (ECCS) flow rate is testimony of clad reaching to temperature lower than its softening temperature during loss of coolant accident (LOCA) in nuclear reactors. Coolant channel(s) of nuclear reactors with vertical fuel-assemblies e. g. LWRs gets voided in a short time as a result of double ended guillotine rupture in coolant pipeline. There is rapid and almost steep rise in clad temperature due to stored energy and decay heat if ECCS flow rate is less than a specified value. A computer program, based on moving mesh methodology, is developed to calculate rewetting velocity. The program is validated using experimental data. Numerical equations are solved by marching technique. The paper will bring out the fact that if coolant flow is less than a specified value the wet front will not reach the top of the clad. This will result some unrewetted clad portion. As the heating of this unrewetted clad is continued it may result softening of clad. If it happens in many channels the integrity of clad as a whole will be lost. There will be high probability that severe accident will take place. The paper presents a method to prevent softening and thus to ensure accident free operation of nuclear reactor. (author)

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

    International Nuclear Information System (INIS)

    Cheong, Jae Hak

    1993-02-01

    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

  7. Severity of vehicle bumper location in vehicle-to-pedestrian impact accidents.

    Science.gov (United States)

    Matsui, Yasuhiro; Hitosugi, Masahito; Mizuno, Koji

    2011-10-10

    Pedestrian protection is one of the key topics for safety measures in traffic accidents all over the world. To analyze the relation between the collision site of the vehicle bumper and the severity of the lower extremity injuries, we performed biomechanical experiments. We compared the applied external force and the risks of subsequent injuries between the impact of the center and side positions of the front bumper. These comparisons were performed by practical impact tests with eight typical different types of cars which were typical of the current vehicle fleets. The tests were made using the TRL legform impactor which was a mechanical substitute of a pedestrian lower extremity. The TRL impactor is used all over the world for assessing the safety of car bumpers. It was found that the risks of lower extremity injuries in the impacts at the side positions, in front of the vehicle's side member, were significantly higher than those at the center. In the tests, we found that foam materials around the rigid front cross member had a significant effect on reducing the lower extremity injury risks and especially tibia fracture risk against vehicle bumper center collisions, but had little effect at the sides of the bumper over the vehicle's side members where the foam was thinner. We also found that the front shape of the vehicle affected the risk of ligaments injuries. According to these results, the information of impact locations of cars in vehicle-to-pedestrian traffic accidents is valuable for clinicians to diagnose patients with lower extremity injuries in traffic accidents and for forensic pathologists to analyze the accident reconstruction. Furthermore, the results suggest that testing of the bumper area in front of the main longitudinal beams should be included in the car safety legislation to require pedestrian safety. Copyright © 2011 Elsevier Ireland Ltd. All rights reserved.

  8. Severe accident management at the Loviisa NPP - Application of integrated ROAAM and PSA level 2

    International Nuclear Information System (INIS)

    Siltanen, S.; Routamo, T.; Tuomisto, H.; Lundstrom, P.

    2007-01-01

    The Risk Oriented Accident Analysis Methodology (ROAAM) was developed for assessment and management of rare, high consequence hazards. The purpose of most ROAAM applications has been to solve major, isolated severe accident issues related to early containment failure such as Mark-I Liner Attack and Direct Containment Heating. In addition to ROAAM in the issue resolution context, the so called Integrated ROAAM approach can be used to provide an overall frame of safety evaluation that allows determination of whether an adequate level of safety has been achieved for a plant. Integrated ROAAM approach brings together quantifications of probabilistic elements based on statistical inference and treatment of deterministic elements based on identification of dominant physics, for severe accident phenomenology, in a well defined and clearly structured way. Fortum, as an owner of the Loviisa NPP, used the Integrated ROAAM approach when developing and implementing a comprehensive severe accident management (SAM) strategy for the Loviisa NPP. The SAM strategy is based on unique features of this VVER-440 plant with ice condenser containment and it includes hardware modifications at the plant, substantial new I and C qualified for severe accident conditions, new SAM guidelines, a SAM Handbook, revision of emergency preparedness organization, and versatile training approaches. It could be argued that the resolution of individual severe accident issues is not sufficient for assessing the overall safety of a nuclear power plant, and thus the ROAAM (in an issue resolution context) is not performing the same function as a PSA study (level 2 included). Actually the Integrated ROAAM approach takes on even a more ambitious task than the PSA, since it determines how a balance can be achieved between accident prevention and mitigation of containment-threatening physical phenomena. Thus it provides a tool for implementing a sound diverse defence-in-depth strategy at a plant. Integrated

  9. Study on entry criteria for severe accident management during hot leg LBLOCAs in a PWR

    International Nuclear Information System (INIS)

    Zhang, Longfei; Zhang, Dafa; Wang, Shaoming

    2007-01-01

    The risk of Large Break Loss of Coolant Accidents (LBLOCA) has been considered an important safety issue since the beginning of the nuclear power industry. The rapid depressurization occurs in the primary coolant circuit when a large break appears in a Pressurized Water Reactors (PWR).Then the coolant temperature reaches saturation at a very low pressure. The core outlet fluid temperatures maybe not reliable indicators of the core damage states at a such lower pressure. The problem is how to decide the time for water injection in the SAM (Severe Accident Management). An alternative entry criterion is the fluid temperature just above the hot channel in which the fluid temperature showed maximum among all the channels. For that reason, a systematic study of entry criterion of SAM for different hot leg break sizes in a 3-loop PWR has been started using the detailed system thermal hydraulic and severe accident analysis code package, RELAP/SCDAPSIM. Best estimate calculations of the large break LOCA of 15 cm, 20 cm and 25 cm without accident managements and in the case of high-pressure safety injection as the accident management were performed in this paper. The analysis results showed that the core exit temperatures are not reliable indicators of the peak core temperatures and core damage states once peak core temperatures reach 1500 K, and the proposed entry criteria for SAM at the time when the core outlet temperature reaches 900 K is not effective to prevent core melt. Then other analyses were performed with a parameter of fluid temperature just above the hot channel. The latter analysis showed that earlier water injection when the fluid temperature just above the hot channel reaches 900 K is effective to prevent further core melt. Since fuel surface and hot channel have spatial distribution and depend on a period of cycle operation, a series of thermocouples are required to install just above the fuel assembly. The maximum exit temperature of 900 K that captured by

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

    International Nuclear Information System (INIS)

    1998-06-01

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

  11. Extension of station blackout coping capability and implications on nuclear safety

    International Nuclear Information System (INIS)

    Volkanovski, Andrija; Prošek, Andrej

    2013-01-01

    Highlights: ► Modifications enhancing station blackout coping capability are analyzed. ► Analysis is done with deterministic and probabilistic safety analysis methods. ► The core heat up is delayed for at least the extension time interval. ► Auxiliary feedwater system delays core heat up even in presence of pumps seal leakage. ► Extension of station blackout coping capability decreases core damage frequency. -- Abstract: The safety of the nuclear power plant depends on the availability of the continuous and reliable sources of electrical energy during all modes of operation of the plant. The station blackout corresponds to a total loss of all alternate current (AC) power as a result of complete failure of both offsite and on-site AC power sources. The electricity for the essential systems during station blackout is provided from the batteries installed in the nuclear power plant. The results of the probabilistic safety assessment show that station blackout is one of the main and frequently the dominant contributor to the core damage frequency. The accident in Fukushima Daiichi nuclear power plants demonstrates the vulnerability of the currently operating nuclear power plants during the extended station blackout events. The objective of this paper is, considering the identified importance of the station blackout initiating event, to assess the implications of the strengthening of the SBO mitigation capability on safety of the NPP. The assessment is done with state-of-art deterministic and probabilistic methods and tolls with application on reference models of nuclear power plants. The U.S. NRC Station Blackout Rule describes procedure for the assessment of the size and capacity of the batteries in the nuclear power plant. The description of the procedure with the application on the reference plant and identified deficiencies in the procedure is presented. The safety analysis is done on reference model of the nuclear power plant. Obtained results show large

  12. Simple probabilistic approach to evaluate radioiodine behavior at severe accidents: application to Phebus test FPT1

    International Nuclear Information System (INIS)

    Rydl, A.

    2007-01-01

    The contribution of radioiodine to risk from a severe accident is recognized to be one of the highest among all the fission products. In a long term (e.g. several days), volatile species of iodine are the most important forms of iodine from the safety point of view. These volatile forms ('volatile iodine') are mainly molecular iodine, I 2 , and various types of organic iodides, RI. A certain controversy exist today among the international research community about the relative importance of the processes leading to volatile iodine formation in containment under severe accident conditions. The amount of knowledge, coming from experiments, of the phenomenology of iodine behavior is enormous and it is embedded in specialized mechanistic or empirical codes. An exhaustive description of the processes governing the iodine behavior in containment is given in reference 1. Yet, all this knowledge is still not enough to resolve some important questions. Moreover, the results of different codes -when applied to relatively simple experiments, such as RTF or CAIMAN - vary widely. Thus, as a complement (or maybe even as an alternative in some instances) to deterministic analyses of iodine behavior, simple probabilistic approach is proposed in this work which could help to see the whole problem in a different perspective. The final goal of using this approach should be the characterization of uncertainties of the description of various processes in question. This would allow for identification of the processes which contribute most significantly to the overall uncertainty of the predictions of iodine volatility in containment. In this work we made a dedicated, small event tree to describe iodine behavior at an accident and we used that tree for a simple sensitivity study. For the evaluation of the tree, the US NRC code EVNTRE was used. To test the proposed probabilistic approach we analyzed results of the integral PHEBUS FPT1 experiment which comprises most of the important

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

    International Nuclear Information System (INIS)

    Van Dorsselaere, Jean-Pierre; Albiol, Thierry; Chaumont, Bernard; Haste, Tim; Journeau, Christophe; Meyer, Leonhard; Sehgal, Bal Raj; Schwinges, Bernd; Beraha, David; Annunziato, Alessandro; Zeyen, Roland

    2010-01-01

    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

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

    International Nuclear Information System (INIS)

    Sammataro, R.F.; Solonick, W.R.; Edwards, N.W.

    1993-01-01

    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)

  15. 47 CFR 76.127 - Satellite sports blackout.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Satellite sports blackout. 76.127 Section 76... Sports Blackout § 76.127 Satellite sports blackout. (a) Upon the request of the holder of the broadcast rights to a sports event, or its agent, no satellite carrier shall retransmit to subscribers within the...

  16. 47 CFR 76.111 - Cable sports blackout.

    Science.gov (United States)

    2010-10-01

    ... 47 Telecommunication 4 2010-10-01 2010-10-01 false Cable sports blackout. 76.111 Section 76.111... CABLE TELEVISION SERVICE Network Non-duplication Protection, Syndicated Exclusivity and Sports Blackout § 76.111 Cable sports blackout. (a) No community unit located in whole or in part within the specified...

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

    International Nuclear Information System (INIS)

    Guentay, S.

    2010-10-01

    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

  18. 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

  19. Research in the Ciemat on severe accidents: strategy and recent results

    International Nuclear Information System (INIS)

    Herranz, L. E.

    2012-01-01

    Severe accident research is a fundamental brick in the nuclear technology wall. Its complexity entails huge challenges that require international cooperation to be overcome. CIEMAT has accumulated more than 40 years of experience in the field. By setting a structured research strategy and a continuous enhancement of theoretical an experimental capabilities, CIEMAT has recently produced the results on which this article builds up. Through them, both its working domains and its firm commitment for a continuous growth of knowledge and know-how are outlined. (Author) 24 refs.

  20. THERMAL AND STRUCTURAL ANALYSIS OF CALANDRIA VESSEL OF A PHWR DURING A SEVERE ACCIDENT

    OpenAIRE

    P.P. KULKARNI; S.V. PRASAD; A.K. NAYAK; P.K. VIJAYAN

    2013-01-01

    In a postulated severe core damage accident in a PHWR, multiple failures of core cooling systems may lead to the collapse of pressure tubes and calandria tubes, which may ultimately relocate inside the calandria vessel forming a terminal debris bed. The debris bed, which may reach high temperatures due to the decay heat, is cooled by the moderator in the calandria. With time, the moderator is evaporated and after some time, a hot dry debris bed is formed. The debris bed transfers heat to the ...

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

    International Nuclear Information System (INIS)

    Higgins, J.C.

    1990-01-01

    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

  2. Numerical study of fundamental processes of severe accidents using a particle method

    International Nuclear Information System (INIS)

    Koshizuka, Seiichi

    2006-01-01

    A particle method has been developed for multiphase flows with large deformation of phase interfaces. The method is called Moving Particle Semi-implicit (MPS) which enables us to analyze incompressible fluid dynamics based on a semi-implicit algorithm. The MPS method has been applied to complex thermal-hydraulic problems in light water reactors and sodium-cooled fast reactors. The present paper provides the review of the past studies using MPS and an introduction of a new research project for severe accident analysis of fast reactors. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Higgins, J.C.

    1990-01-01

    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.

  4. Thermodynamic analysis for the three-layered melt pool during the severe accidents in the APR1400

    International Nuclear Information System (INIS)

    Kang, Kyoungho; Park, Raejoon; Hong, Seongwan

    2009-01-01

    For precise evaluations of the coolability through the External Reactor Vessel Cooling (ERVC) during a severe accident, the melt pool configuration should be accurately defined. The melt pool configurations inside the lower head of the reactor vessel affect the initial thermal load to the vessel and play a key role in determining the integrity of the reactor vessel. In this study, thermodynamic analyses were performed to examine the final melt pool configuration during the severe accidents in the APR1400. As the representative accident scenarios, Large Break Loss of Coolant Accident (LBLOCA), Medium Break Loss of Coolant Accident (MBLOCA), Small Break Loss of Coolant Accident (SBLOCA), Station Black Out (SBO), and Total Loss of Feed Water (TLFW) were selected. The initial melt pool conditions, such as melt mass and melt pool temperature etc., were calculated using the SCDAP/RELAP5/MOD3.3 code. The thermodynamic analyses were performed using the GEMINI code. Combined with the GEMINI code calculations and a peer review of the RASPLAV/MASCA experimental results, the final melt pool configurations were determined for the major accident scenarios of the APR1400. Density evaluation graph was developed for the precise examination of the melt pool layer inversion. The thermodynamic analyses results address the possibility of a melt pool layer inversion in the APR1400 accident sequences. (author)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

    The EPR TM has incorporated several design features in order to specifically address major severe accident safety issues. In particular, it was designed with the objective to transfer high pressure core melt scenarios into a low pressure