WorldWideScience

Sample records for hypothetical severe accident

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

  2. 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.)

  3. Preliminary results of consequence assessment of a hypothetical severe accident using Thai meteorological data

    Science.gov (United States)

    Silva, K.; Lawawirojwong, S.; Promping, J.

    2017-06-01

    Consequence assessment of a hypothetical severe accident is one of the important elements of the risk assessment of a nuclear power plant. It is widely known that the meteorological conditions can significantly influence the outcomes of such assessment, since it determines the results of the calculation of the radionuclide environmental transport. This study aims to assess the impacts of the meteorological conditions to the results of the consequence assessment. The consequence assessment code, OSCAAR, of Japan Atomic Energy Agency (JAEA) is used for the assessment. The results of the consequence assessment using Thai meteorological data are compared with those using Japanese meteorological data. The Thai case has following characteristics. Low wind speed made the radionuclides concentrate at the center comparing to the Japanese case. The squalls induced the peaks in the ground concentration distribution. The evacuated land is larger than the Japanese case though the relocated land is smaller, which is attributed to the concentration of the radionuclides near the release point.

  4. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    International Nuclear Information System (INIS)

    Sarler, B.

    1992-01-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [sl

  5. Study on recriticality of fuel debris during hypothetical severe accidents in the Advanced Neutron Source reactor

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Georgevich, V.; Navarro-Valenti, S.; Shin, S.T.

    1995-09-01

    A study has been performed to measure the potential of recriticality during hypothetical severe accident in Advanced Neutron Source (ANS). For the lumped debris configuration in the Reactor Coolant System (RCS), as found in the previous study, recriticality potential may be very low. However, if fuel debris is dispersed and mixed with heavy water in RCS, recriticality potential has been predicted to be substantial depending on thermal-hydraulic conditions surrounding fuel debris mixture. The recriticality potential in RCS is substantially reduced for the three element core design with 50% enrichment. Also, as observed in the previous study, strong dependencies of k eff on key thermal hydraulic parameters are shown. Light water contamination is shown to provide a positive reactivity, and void formation due to boiling of mixed water provides enough negative reactivity and to bring the system down to subcritical. For criticality potential in the subpile room, the lumped debris configuration does not pose a concern. Dispersed configuration in light water pool of the subpile room is also unlikely to result in criticality. However, if the debris is dispersed in the pool that is mixed with heavy water, the results indicate that a substantial potential exists for the debris to reach the criticality. However, if prompt recriticality disperses the debris completely in the subpile room pool, subsequent recriticality may be prevented since neutron leakage effects become large enough

  6. Modelling of melting and solidification transport phenomena during hypothetical NPP severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Sarler, B [Inst. Jozef Stefan, Ljubljana (Slovenia)

    1992-07-01

    A physical and mathematical framework to deal with the transport phenomena occuring during melting and solidification of the hypothetical NPP severe accidents is presented. It concentrates on the transient temperature, velocity, and species concentration distributions during such events. The framework is based on the Mixture Continuum Formulation of the components and phases, cast in the boundary-domain integral shape structured by the fundamental solution of the Laplace equation. The formulation could cope with various solid-liquid sub-systems through the inclusion of the specific closure relations. The deduced system of boundary-domain integral equations for conservation of mass, energy, momentum, and species could be solved by the boundary element discrete approximative method. (author) [Slovenian] Predstavljeno je fizikalno in matematicno ogrodje za obravnavo prenosnih pojavov taljenja in strjevanja med hipoteticnimi tezkimi nezgodami v jedrskih elektrarnah. Osredotoceno je na popis neustaljene porazdelitve temperatur, hitrosti in koncentracij sestavin med taksnimi dogodki. Ogrodje temelji na formulaciji kontinuuma mesanice komponent in faz, v obliki robno obmocnih integralskih enacb, ki so sestavljena na podlagi fundamentalne resitve Laplace-ove enacbe. Formulacija lahko popisuje stevilne trdno-tekoce pod-sisteme na podlagi specificnih sklopitvenih relacij. Izpeljan sistem robno-obmocnih integralskih enacb za popis ohranitve mase, energije, gibalne kolicine in sestavin lahko resimo na podlagi diskretne aproksimativne metode robnih elementov. (author)

  7. CARNSORE: Hypothetical reactor accident study

    International Nuclear Information System (INIS)

    Walmod-Larsen, O.; Jensen, N.O.; Kristensen, L.; Meide, A.; Nedergaard, K.L.; Nielsen, F.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.

    1984-06-01

    Two types of design-basis accident and a series of hypothetical core-melt accidents to a 600 MWe reactor are described and their consequences assessed. The PLUCON 2 model was used to calculate the consequences which are presented in terms of individual and collective doses, as well as early and late health consequences. The site proposed for the nucelar power station is Carnsore Point, County Wexford, south-east Ireland. The release fractions for the accidents described are those given in WASH-1400. The analyses are based on the resident population as given in the 1979 census and on 20 years of data from the meteorological stations at Rosslare Harbour, 8.5 km north of the site. The consequences of one of the hypothetical core-melt accidents are described in detail in a meteorological parametric study. Likewise the consequences of the worst conceivable combination of situations are described. Finally, the release fraction in one accident is varied and the consequences of a proposed, more probable ''Class 9 accident'' are presented. (author)

  8. Collective radiation doses following a hypothetical, very severe accident to an irradiated fuel transport flask containing AGR fuel

    International Nuclear Information System (INIS)

    Corbett, J.O.

    1985-05-01

    Studies of the consequences of very severe, although unlikely, accidents to irradiated fuel transport flasks are made in order to evaluate risks. If an irradiated fuel transport flask carrying AGR fuel were damaged in a hypothetical accident involving a severe impact followed by a prolonged fire, a small proportion of caesium and other fission products might be released to the atmosphere from the gap inventory of broken fuel pins. The consequent radiation dose to the public would arise predominantly by direct irradiation from ground deposits and the ingestion of slightly contaminated foodstuffs. Although these collective doses must generally be estimated with the aid of computer codes, it is shown here that the worst case, when a high proportion of the radioactivity is deposited in a densely population area, can be assessed approximately by a much simpler method, an approach which is of great value in explaining the calculation in a manner that can be readily understood. A comparison is made between the simple approach and equivalent results from the NECTAR code, the worst case is compared with an ensemble average over all weather conditions, and the relative contributions of the two main routes to collective dose are discussed. (author)

  9. Possibilities of hydrogen removal. Phase 2: Limitation of hydrogen effects in hypothetical severe accidents in PWR reactors

    International Nuclear Information System (INIS)

    Langer, G.; Koehling, A.; Nikodem, H.

    1984-01-01

    In the event of hypothetical severe accidents in light-water reactors, considerable amounts of hydrogen may be produced and released into the containment. Combustion of the hydrogen may jeopardize the integrity of the containment. The study reported here aimed to identify methods to mitigate the hydrogen problem. These methods should either prevent hydrogen combustion, or limit its effects. The following methods have been investigated: pre-inerting; chemical oxygen absorption; removal of oxygen by combustion; post-inerting with N 2 , CO 2 , or halon; aqueous foam; water fog; deliberate ignition; containment purging; and containment venting. The present state of the art in both nuclear and non-nuclear facilities, has been identified. The assessment of the methods was based on accident scenarios assuming significant release of hydrogen and the spectrum of requirements derived from these scenarios was used to determine the advantages and drawbacks of the various methods, assuming their application in a pressurized water reactor of German design. (orig./RW) [de

  10. A study on the overall economic risks of a hypothetical severe accident in nuclear power plant using the delphi method

    International Nuclear Information System (INIS)

    Jang, Han Ki; Kim, Joo Yeon; Lee, Jai Ki

    2008-01-01

    Potential economic impact of a hypothetical severe accident at a nuclear power plant(Uljin units 3/4) was estimated by applying the delphi method, which is based on the expert judgements and opinions, in the process of quantifying uncertain factor. For the purpose of this study, it is assumed that the radioactive plume directs the inland direction. Since the economic risk can be divided into direct costs and indirect effects and more uncertainties are involved in the latter, the direct costs were estimated first and the indirect effects were then estimated by applying a weighting factor to the direct cost. The delphi method however subjects to risk of distortion or discrimination of variables because of the human behavior pattern. A mathematical approach based on the Bayesian inferences was employed for data processing to improve the delphi results. For this task, a model for data processing was developed. One-dimensional Monte Carlo analysis was applied to get a distribution of values of the weighting factor. The mean and median values of the weighting factor for the indirect effects appeared to be 2.59 and 2.08, respectively. These values are higher than the value suggested by OECD/NEA, 1.25. Some factors such as small territory and public attitude sensitive to radiation could affect the judgement of panel. Then the parameters of the model for estimating the direct costs were classified as U- and V-types, and two-dimensional Monte Carlo analysis was applied to quantify the overall economic risk. The resulting median of the overall economic risk was about 3.9% of the Gross Domestic Products (GDP) of Korea in 2006. When the cost of electricity loss, the highest direct cost, was not taken into account, the overall economic risk was reduced to 2.2% of GDP. This assessment can be used as a reference for justifying the radiological emergency planning and preparedness

  11. Predictability of iodine chemistry in the containment of a nuclear power plant under hypothetical severe accident conditions

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Vela-Garcia, M.; Fontanet, J. [Unit of Nuclear Safety Research, CIEMAT, Madrid (Spain)

    2007-07-01

    One of the areas of top interest in the arena of severe accidents to get an accurate prediction of Source Term is Iodine Chemistry. In this paper an assessment of the current capability of MELCOR and ASTEC to predict iodine chemistry within containment in case of a postulated severe accident has been carried out. The experiments FPT1 and FPT2 of the PHEBUS-FP project have been used for comparisons, since they were carried out under rather different containment conditions during the chemistry phase (subcooled vs. saturated sump or acid vs. alkaline pH), which makes them very suitable to assess the current modeling capability of in-containment iodine chemistry models. The results obtained indicate that, even though, both integral codes have specific areas related to iodine chemistry that should be further developed and that their approach to the matter is drastically different, at present ASTEC-IODE allows for a more comprehensive simulation of the containment iodine chemistry. More importantly, lack of maturity of these codes would potentially maximize the so-called user-effect, so that it would be highly recommendable to perform sensitivity studies around iodine chemistry aspects when calculating Source Term scenarios. Key aspects needed of further research are: gaseous iodine chemistry (absent in MELCOR), organic iodine chemistry and adsorption/desorption on/from containment surfaces. (authors)

  12. 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

  13. Nuclear Reactor RA Safety Report, Vol. 16, Maximum hypothetical accident

    International Nuclear Information System (INIS)

    1986-11-01

    Fault tree analysis of the maximum hypothetical accident covers the basic elements: accident initiation, phase development phases - scheme of possible accident flow. Cause of the accident initiation is the break of primary cooling pipe, heavy water system. Loss of primary coolant causes loss of pressure in the primary circuit at the coolant input in the reactor vessel. This initiates safety protection system which should automatically shutdown the reactor. Separate chapters are devoted to: after-heat removal, coolant and moderator loss; accident effects on the reactor core, effects in the reactor building, and release of radioactive wastes [sr

  14. Transition phase in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.; Henninger, R.J.; Jackson, J.F.

    1976-01-01

    Mechanistic analyses of transient-under-cooling accidents have led in some cases to a mild initiating phase instead of a direct hydrodynamic disassembly of the core. The fuel is then trapped in the core by the strong mechanical surroundings and blockages formed by refrozen cladding steel and/or fuel. The formation of fuel blockages has been verified experimentally. The bottled-up core will boil on fission and decay heat, with steel as the working fluid. Boil-up in a churn turbulent flow regime may prevent recriticality due to fuel recompaction. Ultimate fuel removal from the core is probably by a two-phase blow-down after permanent leakage paths are opened. However, a vigorous recriticality can not be precluded. Reactors with void coefficients larger than that in CRBR are more likely to disassemble in the initiating phase, so the transition phase may be unique to small cores

  15. 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.)

  16. Computer codes developed in FRG to analyse hypothetical meltdown accidents

    International Nuclear Information System (INIS)

    Hassmann, K.; Hosemann, J.P.; Koerber, H.; Reineke, H.

    1978-01-01

    It is the purpose of this paper to give the status of all significant computer codes developed in the core melt-down project which is incorporated in the light water reactor safety research program of the Federal Ministry of Research and Technology. For standard pressurized water reactors, results of some computer codes will be presented, describing the course and the duration of the hypothetical core meltdown accident. (author)

  17. The development of a model to study the thermal behaviour of the coolant in the blind elements of a fast sodium-cooled breeder in the case of a severe hypothetical accident during the initial phase

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

    The enthalpy level of the coolant is studied in the interior of gaps and special elements of a fast sodium coded breeder reactor during the initial and the final stages of a hypothetical accident. For this purpose numerical models are presented to calculate the heat transport in the special element on the basis of heat conduction and axial convection. (orig./RW) [de

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

    Energy Technology Data Exchange (ETDEWEB)

    Thoerring, H.; Liland, A.

    2010-12-15

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

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

    International Nuclear Information System (INIS)

    Thoerring, H.; Liland, A.

    2010-12-01

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

  20. Analysis of hypothetical LMFBR whole-core accidents in the USA

    International Nuclear Information System (INIS)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    Methods used for analysis of material behaviour, accident phenomenology and integrated accident calculations are reviewed. Applications of these methods to hypothetical LOF and TOP accidents are discussed. Recent results obtained from applications to FFTF and CRBRP are presented. (author)

  1. Severe accident behavior

    International Nuclear Information System (INIS)

    Denning, R.S.

    1986-01-01

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

  2. 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)

  3. 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

  4. Shock loading of reactor vessel following hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Srinivas, G.; Doshi, J.B.

    1990-01-01

    Hypothetical Core Disruptive Accident (HCDA) has been historically considered as the maximum credible accident in Fast Breeder Reactor systems. Environmental consequences of such an accident depends to a great extent on the ability of the reactor vessel to maintain integrity during the shock loading following an HCDA. In the present paper, a computational model of the reactor core and the surrounding coolant with a free surface is numerical technique. The equations for conservation of mass, momentum and energy along with an equation of state are considered in two dimensional cylindrical geometry. The reactor core at the end of HCDA is taken as a bubble of hot, vaporized fuel at high temperature and pressure, formed at the center of the reactor vessel and expanding against the surrounding liquid sodium coolant. The free surface of sodium at the top of the vessel and the movement of the core bubble-liquid coolant interface are tracked by Marker and Cell (MAC) procedure. The results are obtained for the transient pressure at the vessel wall and also for the loading on the roof plug by the impact of the slug of liquid sodium. The computer code developed is validated against a benchmark experiment chosen to be ISPRA experiment reported in literature. The computer code is next applied to predict the loading on the Indian Prototype Fast Breeder Reactor (PFBR) being developed at Kalpakkam

  5. Intersubassembly incoherencies and grouping techniques in LMFBR hypothetical overpower accident

    International Nuclear Information System (INIS)

    Wilburn, N.P.

    1977-10-01

    A detailed analysis was made of the FTR core using the 100-channel MELT-IIIA code. Results were studied for the transient overpower accident (where 0.5$/sec and 1$/sec ramps) and in which the Damage Parameter and the Failure Potential criteria were used. Using the information obtained from these series of runs, a new method of grouping the subassemblies into channels has been developed. Also, it was demonstrated that a 7-channel representation of the FTR core using this method does an adequate job of representing the behavior during a hypothetical disruptive transient overpower core accident. It has been shown that this new 7-channel grouping method does a better job than an earlier 20-channel grouping. It has also been demonstrated that the incoherency effects between subassemblies as shown during the 76-channel representation of the reactor can be adequately modeled by 7-channels, provided the 7-channels are selected according to the criteria stated in the report. The overall results of power and net reactivity were shown to be only slightly different in the two cases of the 7-channel and the 76-channel runs. Therefore, it can be concluded that any intersubassembly incoherencies can be modeled adequately by a small number of channels, provided the subassemblies making up these channels are selected according to the criteria stated

  6. Severe accident management guidelines

    International Nuclear Information System (INIS)

    Uhle, Jennifer

    2014-01-01

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

  7. Assessment of Loads and Performance of a Containment in a Hypothetical Accident (ALPHA). Facility design report

    International Nuclear Information System (INIS)

    Yamano, Norihiro; Maruyama, Yu; Kudo, Tamotsu; Moriyama, Kiyofumi; Ito, Hideo; Komori, Keiichi; Sonobe, Hisao; Sugimoto, Jun

    1998-06-01

    In the ALPHA (Assessment of Loads and Performance of Containment in Hypothetical Accident) program, several tests have been performed to quantitatively evaluate loads to and performance of a containment vessel during a severe accident of a light water reactor. The ALPHA program focuses on investigating leak behavior through the containment vessel, fuel-coolant interaction, molten core-concrete interaction and FP aerosol behavior, which are generally recognized as significant phenomena considered to occur in the containment. In designing the experimental facility, it was considered to simulate appropriately the phenomena mentioned above, and to cover experimental conditions not covered by previous works involving high pressure and temperature. Experiments from the viewpoint of accident management were also included in the scope. The present report describes design specifications, dimensions, instrumentation of the ALPHA facility based on the specific test objectives and procedures. (author)

  8. 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

  9. EAC european accident code. A modular system of computer programs to simulate LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

    1985-01-01

    One aspect of fast reactor safety analysis consists of calculating the strongly coupled system of physical phenomena which contribute to the reactivity balance in hypothetical whole-core accidents: these phenomena are neutronics, fuel behaviour and heat transfer together with coolant thermohydraulics in single- and two-phase flow. Temperature variations in fuel, coolant and neighbouring structures induce, in fact, thermal reactivity feedbacks which are added up and put in the neutronics calculation to predict the neutron flux and the subsequent heat generation in the reactor. At this point a whole-core analysis code is necessary to examine for any hypothetical transient whether the various feedbacks result effectively in a negative balance, which is the basis condition to ensure stability and safety. The European Accident Code (EAC), developed at the Joint Research Centre of the CEC at Ispra (Italy), fulfills this objective. It is a modular informatics structure (quasi 2-D multichannel approach) aimed at collecting stand-alone computer codes of neutronics, fuel pin mechanics and hydrodynamics, developed both in national laboratories and in the JRC itself. EAC makes these modules interact with each other and produces results for these hypothetical accidents in terms of core damage and total energy release. 10 refs

  10. MCCI study for Pressurized Heavy Water Reactor under hypothetical accident condition

    International Nuclear Information System (INIS)

    Verma, Vishnu; Mukhopadhyay, Deb; Chatterjee, B.; Singh, R.K.; Vaze, K.K.

    2011-01-01

    In case of severe core damage accident in Pressurized Heavy Water Reactor (PHWR), large amount of molten corium is expected to come out into the calandria vault due to failure of calandria vessel. Molten corium at high temperature is sufficient to decompose and ablate concrete. Such attack could fail CV by basement penetration. Since containment is ultimate barrier for activity release. The Molten Core Concrete Interaction (MCCI) of the resulting pool of debris with the concrete has been identified as an important part of the accident sequence. MCCI Analysis has been carried out for PHWR for a hypothetical accident condition where total core material is considered to be relocated in calandria vault. Concrete ablation rate in vertical and radial direction is evaluated for rectangular geometry using MEDICIS module of ASTEC Code. Amount of gases released during MCCI is also evaluated. (author)

  11. SEVERE ACCIDENT MANAGEMENT STATUS AT Loviisa

    International Nuclear Information System (INIS)

    Kymalainen, O.; Tuomisto, H.

    1997-01-01

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

  12. CANDU severe accident analysis

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  13. KADIS: a program to analyse the disassembly phase of hypothetical accidents in LMFBRs

    International Nuclear Information System (INIS)

    Schmuck, P.; Jacobs, G.; Arnecke, G.

    1977-11-01

    The program KADIS models the disassembly phase during power excursions in LMFBR hypothetical accidents. KADIS is based on point kinetics in the neutronics part and on a 2-dimensional representation of the reactor core in the hydrodynamics part. The core is modeled as an ideal, compressible fluid which is heated up adiabatically during the excursion. KADIS was built up with the help of the VENUS program of Argonne National Laboratory. Several important features were added to the basic VENUS model. Therefore we give first a complete description of the mathematical models used. Secondly we provide the user with the necessary information to handle the input/output of KADIS. (orig.) [de

  14. Modeling the consequences of hypothetical accidents for the Titan II system

    International Nuclear Information System (INIS)

    Greenly, G.D.; Sullivan, T.J.

    1981-11-01

    Calculations have been made with the Atmospheric Release Advisory Capability (ARAC) suite of three-dimensional transport and diffusion codes MATHEW/ADPIC to assess the consequences of severe, hypothetical accident scenarios. One set of calculations develops the integrated dose and surface deposition patterns for a non-nuclear, high explosive detonation and dispersal of material. A second set of calculations depicts the time integrated dose and instantaneous concentration patterns for a substantial, continuous leak of the missile fuel oxidizer converted to nitrogen dioxide (NO 2 ). The areas affected and some of the implications for emergency response management are discussed

  15. Simulation and dose analysis of a hypothetical accident in Sanmen nuclear power plant

    International Nuclear Information System (INIS)

    Zhu, Yangmo; Guo, Jianghua; Nie, Chu; Zhou, Youhua

    2014-01-01

    Highlights: • Atmospheric dispersion following a hypothetical accident in Sanmen NPP is simulated. • Japan, North Korea and Russia are slightly influenced in this accident. • In Taiwan and South Korea, population on 100% and 35% of the land should be given information about reducing dose. • In mainland China, about 284 thousand people are likely to get cancer. - Abstract: In November 2013, an AP1000 nuclear power plant (NPP) will be put into commercial operation. An atmospheric dispersion of radionuclides during a severe hypothetical accident in Sanmen NPP, Zhejiang province, China, is simulated with a Lagrangian particle dispersion model FLEXPART. The accident assumes that a station blackout (SBO) accident occurred on August 25, 2011, 55% core was damaged and 49 radionuclides were released into the atmosphere. Our simulation indicates that, during this dispersion, the radioactive plume will cover the mainland China, Taiwan, Japan, North Korea, South Korea and Russia. The radiation dose levels in Japan, North Korea and Russia are the lightest, usually less than 1 mSv. The influenced areas in these countries are 9901 km 2 , 31,736 km 2 and 2,97,524 km 2 , respectively; dose levels in Taiwan and South Korea are moderate, no more than 20 mSv. Information about reducing dose should be given to the public. Total influenced areas in these two countries are 3621 km 2 and 42,370 km 2 , which take up 100% of the land in Taiwan and 35% of the land in South Korea; the worst situation happens in mainland China. The total influenced area is 3 × 106 km 2 and 1,40,000 km 2 in this area has a dose level higher than 20 mSv. Measurement must be taken to reduce the dose. More than 284 thousand residents will face the risk of developing cancer. Furthermore, 96% of this population is mainly concentrated in Zhejiang province, where Sanmen NPP locates

  16. Cernavoda CANDU severe accident evaluation

    International Nuclear Information System (INIS)

    Negut, G.; Marin, A.

    1997-01-01

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

  17. Severe accidents in nuclear reactors

    International Nuclear Information System (INIS)

    Ohai, Dumitru; Dumitrescu, Iulia; Tunaru, Mariana

    2004-01-01

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

  18. 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)

  19. Consequence analyses of hypothetical accidents of high temperature gas-cooled reactors. Pt. 2/3

    International Nuclear Information System (INIS)

    Mueller, A.; Badur, A.

    1978-06-01

    With regard to a hypothetical accident which is characterized by the rupture of the primary circuit and by the additional failure of active engineered safeguards, the fission product release resulting from the unlimited core heat-up is analyzed. The applied models are explained and the data base being used is documented. The generally conservative treatment yields pessimistic activity release rates into the containment. The results show in particular that spontaneous massive fission product release does not occur. The time-dependency of the activity release from the fuel elements, the primary circuit and at last from the containment leads to a time delay in the range of at least several hours, before the environmental radiation load is raised. Ultimately the maximum radiation load itself proves relatively favourable. (orig.) 891 HP [de

  20. Risks and consequences of a hypothetical radiological accident on nuclear powered submarine traversing Suez canal

    International Nuclear Information System (INIS)

    Salama, Mohamed

    2008-01-01

    Full text: Egypt has unique problem in Suez Canal, although there are, a number of radioactive Cargos traveling through the Canal which includes new and spent reactor fuel and about 100 metric tons of uranium hexafluoride each year, under the regulatory control of the Egyptian Atomic Energy Authority, there is, still a major problem concerning the passage of a number of nuclear powered vessels and submarines passing through the canal several times each year. The passage of these vessels and submarines has a political situation and not under the regulatory control of the Egyptian regulatory body. In spite of all precautions that are taken, in the nuclear powered vessels and submarines from the point of view of the rugged design of the reactor plant, multiple safety systems and operation with exceptional consideration for safety. Although of all of these a potential for a serious accident may does arise, even though, its probability is minimal. The Government of Egypt has established a national radiological emergency plan in order to cope with any radiological accidents, which may arise inside the country. Suez Canal lies in the north east of Egypt, and passes through a zone of considerable business, agriculture and industrial activities. The zone consists of three populated provinces, Port Said, Ismailia and Suez. According to Suez Canal authority regulations it is not allowed for these vessels and submarines to be landed in port. The motivation of the present paper was undertaken to discuss a hypothetical nuclear reactor accident aboard a nuclear powered submarine occurred during its passage in the Suez Canal. Such an accident will produce a radioactive cloud containing a number of radioactive materials. In such type of accidents contamination and causality zones, could extend to several kilometers. The different phases of the accident are going to be discussed and analyzed. The emergency actions taken during the accident phases are going to be presented. The

  1. Effects of spent fuel types on offsite consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Courtney, J. C.; Dwight, C. C.; Lehto, M. A.

    2000-01-01

    Argonne National Laboratory (ANL) conducts experimental work on the development of waste forms suitable for several types of spent fuel at its facility on the Idaho National Engineering and Environmental Laboratory (INEEL) located 48 km West of Idaho Falls, ID. The objective of this paper is to compare the offsite radiological consequences of hypothetical accidents involving the various types of spent nuclear fuel handled in nonreactor nuclear facilities. The highest offsite total effective dose equivalents (TEDEs) are estimated at a receptor located about 5 km SSE of ANL facilities. Criticality safety considerations limit the amount of enriched uranium and plutonium that could be at risk in any given scenario. Heat generated by decay of fission products and actinides does not limit the masses of spent fuel within any given operation because the minimum time elapsed since fissions occurred in any form is at least five years. At cooling times of this magnitude, fewer than ten radionuclides account for 99% of the projected TEDE at offsite receptors for any credible accident. Elimination of all but the most important nuclides allows rapid assessments of offsite doses with little loss of accuracy. Since the ARF (airborne release fraction), RF (respirable fraction), LPF (leak path fraction) and atmospheric dilution factor (χ/Q) can vary by orders of magnitude, it is not productive to consider nuclides that contribute less than a few percent of the total dose. Therefore, only 134 Cs, 137 Cs- 137m Ba, and the actinides significantly influence the offsite radiological consequences of severe accidents. Even using highly conservative assumptions in estimating radiological consequences, they remain well below current Department of Energy guidelines for highly unlikely accidents

  2. Assessment of environmental public exposure from a hypothetical nuclear accident for Unit-1 Bushehr nuclear power plant

    Energy Technology Data Exchange (ETDEWEB)

    Sohrabi, M.; Ghasemi, M.; Amrollahi, R.; Khamooshi, C.; Parsouzi, Z. [Amirkabir University of Technology, Health Physics and Dosimetry Research Laboratory, Department of Physics, Tehran (Iran, Islamic Republic of)

    2013-05-15

    Unit-1 of the Bushehr nuclear power plant (BNPP-1) is a VVER-type reactor with 1,000-MWe power constructed near Bushehr city at the coast of the Persian Gulf, Iran. The reactor has been recently operational to near its full power. The radiological impact of nuclear power plant (NPP) accidents is of public concern, and the assessment of radiological consequences of any hypothetical nuclear accident on public exposure is vital. The hypothetical accident scenario considered in this paper is a design-basis accident, that is, a primary coolant leakage to the secondary circuit. This scenario was selected in order to compare and verify the results obtained in the present paper with those reported in the Final Safety Analysis Report (FSAR 2007) of the BNPP-1 and to develop a well-proven methodology that can be used to study other and more severe hypothetical accident scenarios for this reactor. In the present study, the version 2.01 of the PC COSYMA code was applied. In the early phase of the accidental releases, effective doses (from external and internal exposures) as well as individual and collective doses (due to the late phase of accidental releases) were evaluated. The surrounding area of the BNPP-1 within a radius of 80 km was subdivided into seven concentric rings and 16 sectors, and distribution of population and agricultural products was calculated for this grid. The results show that during the first year following the modeled hypothetical accident, the effective doses do not exceed the limit of 5 mSv, for the considered distances from the BNPP-1. The results obtained in this study are in good agreement with those in the FSAR-2007 report. The agreement obtained is in light of many inherent uncertainties and variables existing in the two modeling procedures applied and proves that the methodology applied here can also be used to model other severe hypothetical accident scenarios of the BNPP-1 such as a small and large break in the reactor coolant system as well

  3. Comprehensive and consistent interpretation of local fault experiments and application to hypothetical local overpower accident in Monju

    International Nuclear Information System (INIS)

    Fukano, Yoshitaka

    2013-01-01

    Experimental studies on local fault (LF) accidents in fast breeder reactors have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Comprehensive and consistent interpretations of in-pile and out-of-pile experiments related to LF were arrived at in this study based on state-of-the-art review and data analysis techniques. Safety margins for a hypothetical local overpower accident, which was evaluated as a LF accident in the licensing document of the construction permit for a prototype fast breeder reactor called Monju, were also studied. Based on comprehensive interpretations of the latest experimental database, including those performed after the permission of Monju construction, it was clarified that the evaluation of the hypothetical local overpower accident in the Monju licensing was sufficiently conservative. Furthermore, it incorporated adequate safety margins in terms of failure thresholds of the fuel pin, molten fuel ejection, fuel sweep-out behavior after molten fuel ejection, and pin-to-pin failure propagation. Moreover, these comprehensive interpretations are valid and applicable to the safety evaluation of LF accidents of other fast breeder reactors with various fuel and core designs. (author)

  4. ATHENA-2D: A computer code for simulation of hypothetical recriticality accidents in a thermal neutron spectrum

    International Nuclear Information System (INIS)

    Schwinkendorf, K.N.

    1995-01-01

    In a damaged light water reactor core (as in the aftermath of a Three-Mile-Island-like core meltdown), water reflood is performed to carry off decay heat. The severely degraded geometry of the fuel debris bed may increase core reactivity with water reflood. Sufficient boron poisoning of the reflood water is therefore very important. One hypothetical accident is the reintroduction of cooling water that is insufficiently borated, resulting in the damaged reactor attaining criticality in this uncontrolled configuration. The goal in simulating this accident is the prediction of the energy release from the resulting transient

  5. Calculation of individual and population doses on Danish territory resulting from hypothetical core-melt accidents at the Barsebaeck reactor

    International Nuclear Information System (INIS)

    1977-01-01

    Individual and population doses within Danish territory are calculated from hypothetical, severe core-melt accidents at the Swedish nuclear plant at Barsebaeck. The fission product inventory of the Barsebaeck reactor is calculated. The release fractions for the accidents are taken from WASH-1400. Based on parametric studies, doses are calculated for very unfavourable, but not incredible weather conditions. The probability of such conditions in combination with wind direction towards Danish territory is estimated. Doses to bone marrow, lungs, GI-tract and thyroid are calculated based on dose models developed at Risoe. These doses are found to be consistent with doses calculated with the models used in WASH-1400. (author)

  6. Iodine chemistry at severe accidents. A review and evaluation of the state-of-the-art in the field. APRI 5 report. Part I: Iodine chemistry at hypothetical severe accidents. A review of the state-of-the-art 2003. Part II: A comparison of our knowledge on iodine chemistry and fission products with the current models used in MAAP 4.0.5

    International Nuclear Information System (INIS)

    Liljenzin, Jan-Olov

    2005-01-01

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

  7. Prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

    Weisshaeupl, H.

    1996-01-01

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

  8. Comparison of SAS3A and MELT-III predictions for a transient overpower hypothetical accident

    International Nuclear Information System (INIS)

    Wilburn, N.P.

    1976-01-01

    A comparison is made of the predictions of the two major codes SAS3A and MELT-III for the hypothetical unprotected transient overpower accident in the FFTF. The predictions of temperatures, fuel restructuring, fuel melting, reactivity feedbacks, and core power are compared

  9. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  10. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

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

    2001-01-01

    with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality-both super-prompt power bursts and quasi steady-state power......Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies......, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g(-1), was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s(-1). In most cases, however, the predicted energy deposition was smaller, below...

  11. Hypothetical accidents of light-water moderated nuclear power plants in the framework of emergency planning

    International Nuclear Information System (INIS)

    1979-07-01

    Hypothetical accidents in nuclear power plants are events which by definition can have a devastating impact on the surroundings of the plant. Apart from an adequate plant design, the protection of the population in case of an accident is covered by the emergency planning. Of major importance are the measures for the short-term emergency protection. The decision on whether these measures are applied has to be based on appropriate measurements within the plant. The aim and achieved result of this investigation is to specify accident types. They serve as operational decision making criteria to determine the necessary measurements for analysing the accident in the accident situation, and to provide indications for choosing the suitable strategy for the protection measures. (orig.) [de

  12. 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

  13. An approach for estimating the radiological significance of a hypothetical major nuclear accident over long distance transboundary scales

    Energy Technology Data Exchange (ETDEWEB)

    Mitrakos, D., E-mail: dimitris.mitrakos@eeae.gr; Potiriadis, C.; Housiadas, C.

    2016-04-15

    Highlights: • Actions may be warranted after a major nuclear accident even at long distances. • Distance may not be the decisive parameter for longer term radiological impact. • Remote impact may vary orders of magnitude depending on the meteorological conditions. • The potential impact can be assessed using computationally inexpensive calculations. - Abstract: After the Fukushima accident important initiatives were taken in European level to enhance the nuclear safety level of the existing and planned nuclear reactors, such as the so-called nuclear “stress-tests” and the amendment of the Nuclear Safety Directive. A recent work of HERCA and WENRA focused on the need for a more consistent and harmonized response in a transboundary context in case of a hypothetical major nuclear accident in Europe. Such an accident, although very improbable, cannot be totally excluded and so, should be considered in emergency preparedness arrangements among the various European countries. In case of a hypothetical severe Fukushima-like accident in Europe, the role of the neighboring countries may be important, since the authorities should be able to provide information and advice to the government and the public, but also can contribute to the overall assessment of the situation be their own means. In this work we assess the radiological significance of a hypothetical major nuclear accident for distances longer than 300 km that are not typically covered by the internationally accepted emergency planning zones. The approach is simple and computationally inexpensive, since it is based on the calculation of only a few release scenarios at dates selected within a whole year on the basis of bounding the deposition levels at long distances in relation to the occurrence of precipitation. From the calculated results it is evident that distance is not the only decisive parameter in estimating the potential radiological significance of a severe nuclear accident. The hypothetical

  14. Extension of emergency operating procedures for severe accident management

    International Nuclear Information System (INIS)

    Chiang, S.C.

    1992-01-01

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

  15. A CANDU Severe Accident Analysis

    International Nuclear Information System (INIS)

    Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

    2006-01-01

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

  16. Severe accident source term reassessment

    International Nuclear Information System (INIS)

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

    1987-01-01

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

  17. Radiological consequences of a hypothetical ''roof breakdown'' accident of the Chernobyl sarcophagus

    International Nuclear Information System (INIS)

    Pretzsch, G.

    1997-01-01

    On behalf of the German Federal Ministry for Environment, Nature Conservation and Nuclear Safety GRS performed investigations with the aim to improve the safety of the Chernobyl Unit 4 shelter in close connection with the Ministry for Environment and Nuclear Safety of the Ukraina from 1992 to 1995. One of the tasks of the working programme was concerned with the analysis of hypothetical accidents of the present shelter, which comprises the newly built Sarcophagus and the remaining ruins of Unit 4. In close collaboration with Ukrainian and Russian experts the maximum hypothetical accident was defined to be the breakdown of the roof of the Sarcophagus and subsequent release of the radioactive dust which is mainly located in the destroyed reactor hall and the neighboring rooms

  18. Consequences in Norway of a hypothetical accident at Sellafield: Potential release - transport and fallout

    International Nuclear Information System (INIS)

    Ytre-Eide, M. A.; Standring, W.J.F.; Amundsen, I.; Sickel, M.; Liland, A.; Saltbones, J.; Bartnicki, J.; Haakenstad, H.; Salbu, B.

    2009-03-01

    This report focuses on transport and fallout from 'worst-case' scenarios based on a hypothetical accident at the B215 facility for storing Highly Active Liquors (HAL) at Sellafield. The scenarios involve an atmospheric release of between 0.1-10 % of the total HAL inventory; only transport and fallout of 137 Cs is considered in this case study. Simulations resulted in between 0.1-50 times the maximum 137 Cs fallout experienced in the most contaminated areas in Norway after the Chernobyl accident. (Author)

  19. Analysis of hypothetical LMFBR whole-core accidents in the USA

    International Nuclear Information System (INIS)

    Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

    1978-01-01

    The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

  20. 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.

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

    International Nuclear Information System (INIS)

    Thoerring, H.; Ytre-Eide, M.A.; Liland, A.

    2010-12-01

    This report describes the possible environmental consequences for Norway due to a hypothetical accident at the Sellafield complex in the UK. The scenario considered involves an explosion and fire at the B215 facility resulting in a 1 % release of the total HAL (Highly Active liquor) inventory of radioactive waste with a subsequent air transport and deposition in Norway. Air transport modelling is based on real meteorological data from October 2008 with wind direction towards Norway and heavy precipitation. This weather is considered to be quite representative as typical seasonal weather. Based on this weather scenario, the estimated fallout in Norway will be ∼ 17 P Bq of caesium-137 which is 7 times higher than the fallout from the Chernobyl accident. The modelled radioactive contamination is linked with data on transfer to the food chain and statistics on production and hunting to assess the consequences for foodstuffs. The investigation has been limited to the terrestrial environment, focussing on wild berries, fungi, and animals grazing unimproved pastures (i.e. various types of game, reindeer, sheep and goats). The predicted consequences are severe - especially in connection to sheep and goat production. Up to 80 % of the lambs in Norway could be exceeding the food intervention levels for radiocaesium the first years after the fallout, with 30-40 % likely to be above for many years. There will, consequently, be a need for extensive countermeasures in large areas for years or even decades involving several hundred thousand animals each year. Large consequences are also expected for reindeer husbandry - the first year in particular due to the time of fallout which is just prior to winter slaughter. The consequences will be most sever for reindeer herding in middle and southern parts of Norway, but problems may reach as far north as Finnmark where we find the majority of Norwegian reindeer production. The consequences for game will mostly depend on the regional

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

    Energy Technology Data Exchange (ETDEWEB)

    Thoerring, H.; Ytre-Eide, M.A.; Liland, A.

    2010-12-15

    This report describes the possible environmental consequences for Norway due to a hypothetical accident at the Sellafield complex in the UK. The scenario considered involves an explosion and fire at the B215 facility resulting in a 1 % release of the total HAL (Highly Active liquor) inventory of radioactive waste with a subsequent air transport and deposition in Norway. Air transport modelling is based on real meteorological data from October 2008 with wind direction towards Norway and heavy precipitation. This weather is considered to be quite representative as typical seasonal weather. Based on this weather scenario, the estimated fallout in Norway will be approx 17 P Bq of caesium-137 which is 7 times higher than the fallout from the Chernobyl accident. The modelled radioactive contamination is linked with data on transfer to the food chain and statistics on production and hunting to assess the consequences for foodstuffs. The investigation has been limited to the terrestrial environment, focussing on wild berries, fungi, and animals grazing unimproved pastures (i.e. various types of game, reindeer, sheep and goats). The predicted consequences are severe - especially in connection to sheep and goat production. Up to 80 % of the lambs in Norway could be exceeding the food intervention levels for radiocaesium the first years after the fallout, with 30-40 % likely to be above for many years. There will, consequently, be a need for extensive countermeasures in large areas for years or even decades involving several hundred thousand animals each year. Large consequences are also expected for reindeer husbandry - the first year in particular due to the time of fallout which is just prior to winter slaughter. The consequences will be most sever for reindeer herding in middle and southern parts of Norway, but problems may reach as far north as Finnmark where we find the majority of Norwegian reindeer production. The consequences for game will mostly depend on the

  3. Effect of engineered safety features on the risk of hypothetical LMFBR accidents

    International Nuclear Information System (INIS)

    Cybulskis, P.

    1978-01-01

    The risks of hypothetical core-disruptive accidents in liquid-metal-cooled fast breeder reactors which involve meltthrough of the reactor vessel are compared for two plant designs: one design without specific provisions to accommodate such an accident and the other design with an ex-vessel core catcher and a cvity hot liner. The approach to risk analysis used is that developed in the Reactor Safety Study (WASH-1400). Since the probability of occurrence of such an event has not been evaluated, however, insight into the potential risk is gained only on a relative basis. The principal conclusions of this study are: (1) adding a core catcher--hot liner reduces the probabilty of accidents having major consequences; (2) the degree to which hot liner--core catcher systems can reduce the risk of melt-through accidents is limited by the failure probability of these systems; (3) fractional radioactive releases to the environment in the liquid-metal-cooled fast breeder reactor accidents considered are comparable to those from the light-water reactors evaluated in WASH-1400; (4) since sodium--concrete reactions are a dominant driving force during the accident, the integrity of the cavity liner is as important as the function of the core catcher; (5) there may be other accidents or paths to radioactive releases that are not affected by the addition of a hot liner--core catcher

  4. Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

    International Nuclear Information System (INIS)

    Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

    1982-01-01

    A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

  5. Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Sung-yeop; Lim, Ho-Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study.

  6. Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

    International Nuclear Information System (INIS)

    Kim, Sung-yeop; Lim, Ho-Gon

    2016-01-01

    Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study

  7. 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

  8. 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

  9. Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Albrecht, H.; Malinauskas, A.P.

    1978-01-01

    A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

  10. 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

  11. CANDU safety under severe accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  12. Guide to General Atomic studies of hypothetical nuclear driven accidents for the Fort St. Vrain reactor

    International Nuclear Information System (INIS)

    Wei, T.; Tobias, M.

    1974-03-01

    The work of the General Atomic Company (GAC) in preparing those portions of the Final Safety Analysis Report for the Fort St. Vrain Reactor (FSV) having to do with hypothetical nuclear driven accidents has been reviewed and a guide to this literature has been prepared. The sources for this study are the Final Safety Analysis Report itself, the Quarterly and Monthly Progress Reports, Topical Reports, and Technical Specifications. The problems considered and the methods used are outlined. An appendix gives a systematic analysis which was used as a guide in organizing the references. (U.S.)

  13. Models and methods for predicting the release of fission products during hypothetical accidents in HTGRs

    International Nuclear Information System (INIS)

    Bailly, H.W.

    1988-01-01

    The paper deals with experiments, computational models and methods used to describe the fission product transport (diffusion and particle failure) in the fuel elements of a pebble-bed high-temperature module reactor (HTGR Module) during hypothetical accidents. The codes which describe the diffusion of fission products in the fuel elements are e.g. GETTER and FRESCO. PANAMA, IA/KWU failure function and the so called GOODIN models describe the particle failure. All these models may be used in the risk analysis. The experimental results obtained at the Nuclear Research Center Julich, Germany are discussed and compared with the model calculations for these experiments

  14. Study of an hypothetical reactor meltdown accident for a 50 MW sub(th) fast reactor

    International Nuclear Information System (INIS)

    Azevedo, E.M. de.

    1983-01-01

    A melhodology for determining the energy released in hypothetical reactor meltdown accidents is presented. A numerical code was developed based upon the Nicholson method for a uniform and homogeneous reactor with spherical geometry. A comparative study with other know programs in the literature which use better approximations for small energy released, shows that the methodology used were compatible with those under comparison. Besides the influence of some parameters on the energy released, such as the initial power level and the prompt neutron lifetime was studied under this metodology and its result exhibitted. The Doppler effect was also analyzed and its influence on the energy released has been emphasized. (Author) [pt

  15. 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

  16. 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

  17. Severe accident recriticality analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-11-01

    Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s{sup -1} injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g{sup -1}, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s{sup -1}. In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated

  18. Severe accident recriticality analyses (SARA)

    International Nuclear Information System (INIS)

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

    2001-01-01

    Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the following issues have been investigated in the SARA project: (1) the energy deposition in the fuel during super-prompt power burst; (2) the quasi steady-state reactor power following the initial power burst; and (3) containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality--both super-prompt power bursts and quasi steady-state power generation--for the range of parameters studied, i.e. with core uncovering and heat-up to maximum core temperatures of approximately 1800 K, and water flow rates of 45-2000 kg s -1 injected into the downcomer. Since recriticality takes place in a small fraction of the core, the power densities are high, which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal g -1 , was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate of 2000 kg s -1 . In most cases, however, the predicted energy deposition was smaller, below the regulatory limits for fuel failure, but close to or above recently observed thresholds for fragmentation and dispersion of high burn-up fuel. The highest calculated quasi steady

  19. Severe Accident Recriticality Analyses (SARA)

    International Nuclear Information System (INIS)

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

    1999-11-01

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

  20. The role of fission gas in the analysis of hypothetical core disruptive accidents

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, E A [Gesellschaft fuer Kernforschung mbH, INR Kernforschungszentrum, Karlsruhe (Germany)

    1977-07-01

    This paper summarizes recent work at Karlsruhe with the goal of understanding the effects of fission gas in hypothetical core disruptive accidents. The fission gas behavior model is discussed. The computer programs LANGZEIT and KURZZEIT describe the long-term and the transient gas behavior, respectively. Recent improvements in the modeling and a comparison of results with experimental data are reported. A somewhat detailed study of the role of fission gas in transient overpower (TOP) accidents was carried out. If pessimistic assumptions, like pin failure near the axial midplane are made, these accidents end in core disassembly. The codes HOPE and KADIS were used to analyze the initiating and the disassembly phase in these studies. Improvements of the codes are discussed. They include an automatic data transfer from HOPE to KADIS, and a new equation of state in KADIS, with an improved model for fission gas behavior. The analysis of a 15 cents/sec reactivity ramp accident is presented. Different pin failure criteria are used. In the cases selected, the codes predict an energetic disassembly. For the much discussed loss-of-flow driven TOP, detailed models are presently not available at Karlsruhe. Therefore, only a few comments and the results of a few scoping calculations will be presented.

  1. 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

  2. Convective behaviour in severe accidents

    International Nuclear Information System (INIS)

    Clement, C.F.

    1988-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

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

    International Nuclear Information System (INIS)

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

    1997-01-01

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

  6. Severe accidents: in nuclear power plants

    International Nuclear Information System (INIS)

    1986-01-01

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

  7. SAMEX: A severe accident management support expert

    International Nuclear Information System (INIS)

    Park, Soo-Yong; Ahn, Kwang-Il

    2010-01-01

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

  8. Simulation of a hypothetical core disruptive accident in the mars test-facility

    International Nuclear Information System (INIS)

    Robbe, M.F.; Lepareux, M.

    2001-01-01

    In France, a large experimental programme MARA/MARS was undertaken in the 80's to estimate the mechanical consequences of an HCDA (Hypothetical Core Disruptive Accident) and to validate the SIRIUS computer code used at that time for the numerical simulations. At the end of the 80's, it was preferred to add a HCDA sodium-bubble-argon tri-component constitutive law to the general ALE fast dynamics finite element CASTEM-PLEXUS code rather than going on developing and using the specialized SIRIUS code. The experimental results of the MARA programme were used in the 90's to validate and qualify the CASTEM-PLEXUS code. A first series of computations of the tests MARA 8, MARA 10 and MARS was realised. The simulations showed a rather good agreement between the experimental and computed results for the MARA 8 and MARA 10 tests - even if there were some discrepancies - but the prediction of the MARS structure displacements and strains was overestimated. This conservatism was supposed to come from the fact that several MARS non axisymmetric structures like core elements, pumps and heat exchangers were not represented in the CASTEM-PLEXUS model. These structures, acting as porous barriers, had a protective effect on the mock-up containment by absorbing energy and slowing down the fluid impacting the containment. For these reasons, we developed in CASTEM-PLEXUS a new HCDA constitutive law taking into account the presence of the internal structures (without meshing them) by means of an equivalent porosity method. In other respects, the process used for dealing with the fluid-structure coupling in CASTEM-PLEXUS was improved. Thus a second series of simulations of the tests MARA8 and MARA10 was realised. A simulation of the test MARS was carried out too with the same simplified representation of the peripheral structures as in order to estimate the improvement provided by the new fluid-structure coupling. This paper presents a third numerical simulation of the MARS test with the

  9. 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

  10. Radiation dose evaluation for hypothetical accident with transport package containing Iridium-192 source

    International Nuclear Information System (INIS)

    Trontl, K.; Bace, M.; Pevec, D.

    2002-01-01

    The aim of this paper is to evaluate dose rates for a hypothetical accident with transport package containing Iridium-192 source and to design additional shielding necessary for the safe unloading of the container, assuming that during the unloading process the whole contents of a radioactive source is unshielded and that the operation is going to take place at the site where a working area exists in the vicinity of the unloading location. Based on the calculated radiation dose rates, a single arrangement of the additional concrete shields necessary for reduction of the gamma dose rates to the permitted level is proposed. The proposed solution is optimal considering safety on one hand and costs on the other.(author)

  11. An analysis of reactor structural response to fuel sodium interaction in a hypothetical core disruptive accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A., calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. This work was supported by a grant from Power Reactor and Nuclear Fuel Development Corporation. (auth.)

  12. 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

  13. NPP Krsko Severe Accident Management Guidelines Implementation

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  14. Simulation of thermal response of the 250 MWT modular HTGR during hypothetical uncontrolled heatup accidents

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.

    1985-01-01

    One of the central design features of the 250 MWT modular HTGR is the ability to withstand uncontrolled heatup accidents without severe consequences. This paper describes calculational studies, conducted to test this design feature. A multi-node thermal-hydraulic model of the 250 MWT modular HTGR reactor core was developed and implemented in the IBM CSMP (Continuous System Modeling Program) simulation language. Survey calculations show that the loss of forced circulation accident with loss of steam generator cooling water and with accidental depressurization is the most severe heatup accident. The peak hot-spot fuel temperature is in the neighborhood of 1600 0 C. Fuel failure and fission product releases for such accidents would be minor. Sensitivity studies show that code input assumptions for thermal properties such as the side reflector conductivity have a significant effect on the peak temperature. A computer model of the reactor vessel cavity concrete wall and its surrounding earth was developed to simulate the extremely unlikely and very slowly-developing heatup accident that would take place if the worst-case loss of forced primary coolant circulation accident were further compounded by the loss of cooling water to the reactor vessel cavity liner cooling system. Results show that the ability of the earth surrounding the cavity to act as a satisfactory long-term heat sink is very sensitive to the assumed rate of decay heat generation and on the effective thermal conductivity of the earth

  15. Severe accident training simulator APROS SA

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  16. Release of gases and their influence on containment integrity during a hypothetical meltdown accident

    International Nuclear Information System (INIS)

    Hassmann, K.; Reimann, M.

    1981-01-01

    The sequence of a hypothetical core melt down accident has been subdivided into four phases. Heating up of the core until failure of the core support structure is the first phase. It starts at a certain water level in the reactor pressure vessel (RPV) and ends with the failure of the grid plate. The second phase is characterized by the evaporation of the water in the lower plenum of the RPV. The second phase lasts until a molten core debris is formed. The third phase is concerned with the heating up of the pressure vessel after formation of a molten pool in the lower plenum of the RPV. After pressure vessel failure, the molten corium will interact in the fourth phase with the concrete structure beneath the pressure vessel. In this paper the gas release during all four accident phases and the resulting pressure-time history within the containment of a German standard PWR is given, taking into account violent combustion of hydrogen. In particular, the differences caused by dsestruction of concrete with silicious and with calcareous aggregates has been analyzed. The basis for the results in the 4th phase is the WECHSL code. Long term containment calculations have been performed with the COCMEL-code

  17. 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

  18. 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.

  19. Monitoring severe accidents using AI techniques

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-05-15

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

  20. National practices in relation to severe accidents

    International Nuclear Information System (INIS)

    Soda, Kunihisa

    1989-01-01

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

  1. Iodine behaviour in severe accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  2. 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

  3. Chemical phenomena under severe accident conditions

    International Nuclear Information System (INIS)

    Powers, D.A.

    1988-01-01

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

  4. Radiation doses estimation for hypothetical NPP Krsko accidents without and with PCFV using RASCAL software

    International Nuclear Information System (INIS)

    Vukovic, Josip; Konjarek, Damir; Grgic, Davor

    2014-01-01

    Calculation is done using Source Term to Dose module of RASCAL (Radiological Assessment System Consequence Analysis) software to estimate projected radiation doses from a radioactive plume to the environment. Utilizing this module, it is possible to do preliminary assessment of consequences to the environment in case of adverse reactor conditions or releases from other objects containing radioactive materials before the emergency situation has happened or in the early phase. RASCAL is simple, easy to use, fast-running tool able to provide initial estimate of radiological consequences of nuclear accidents. Upon entering rather limited amount of input parameters for the Krsko NPP, mostly key plant parameters, time dependent source term calculation is executed to determine radioactive inventory release rates for different plant conditions, release paths and availability of protective measures. These rates given for each radionuclide as a function of time are used as an input to atmospheric dispersion and transport model. Together with release rates, meteorological conditions dataset serve as input to determine the behavior of the radioactive releases that is plume in the atmosphere. So as an output, RASCAL produces a 'dispersion envelope' of radionuclide concentrations in the atmosphere. These concentrations of radionuclides in the atmosphere are further used for estimating the doses to the environment and the public downwind the release point. Throughout this paper, dose assessment is performed for two distances, close-in distance and distance out to 40 km from the source, for hypothetical NPP Krsko accidents without and with Passive Containment Filtered Vent (PCFV) system used. Obvious difference is related to released radioactivity of Iodine isotopes. Results of radioactive effluents deposition in the environment are displayed via various doze parameters, radionuclide concentrations and materials exposure rates in this particular case. (authors)

  5. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  6. 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)

  7. 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)

  8. Determination of the maximum individual dose exposure resulting from a hypothetical LEU plate-melt accident

    International Nuclear Information System (INIS)

    Abdelhady, Amr

    2013-01-01

    Highlights: ► Studying the radioactive release results from hypothetical plate-melt accident. ► Hotspot code was used to study the dose distributions around the reactor. ► A 90% decrease in the received dose in proper operation of filtration. ► The received dose is lower than the annual permissible dose after filtration. - Abstract: The objective of this study was to provide an estimate of the potential impact of accidental radioactive release from the testing cell of the Egyptian second research reactor ETRR-2 on the dose level of public around the reactor. The assessment was performed for two cases: an evaluation of the impact that accidental release has on the dose that would be received by public around the reactor in case of proper operation of testing cell filtration system; and an assessment of the potential dose in case of loss of testing cell filtration system. The results show that the filtration system has a great role in decreasing the dose received by an individual located outside the reactor to a dose level lower than the annual permissible dose

  9. Simulation of hypothetical criticality accidents involving homogeneous damp low-enriched UO2 powder systems

    International Nuclear Information System (INIS)

    Basoglu, B.; Brewer, R.W.; Haught, C.F.; Hollenbach, D.F.; Wilkinson, A.D.; Dodds, H.L.; Pasqua, P.F.

    1994-01-01

    This paper describes the development of a computer model for predicting the excursion characteristics of a postulated, hypothetical, critically accident involving a homogeneous mixture of low-enriched UO 2 powder and water contained in a cylindrical blender. The model uses point neutronics coupled with simple lumped-parameter thermal-hydraulic feedback. The temperature of the system is calculated using a simple time-dependent energy balance where two extreme conditions for the thermal behavior of the system are considered, which bound the real life situation. Using these extremes, three different models are developed. To evaluate the models, the authors compared the results with the results of the POWDER code, which was developed by the Commissariat a l'Energie Atomique/United Kingdom Atomic Energy Authority (CEA/UKAEA) for damp powder systems. The agreement in these comparisons is satisfactory. Results of the excursion studies in this work show that approximately 10 19 fissions occur as a result of accidental water ingress into powder blenders containing 5,000 kg of low-enriched (5%) UO 2 powder

  10. Safety assessment of the Indonesian multipurpose reactor RSG-GAS against ATWS and hypothetical accidents

    International Nuclear Information System (INIS)

    Hastowo, H.; Nabbi, R.; Prayoto; Ismuntoyo, R.P.H.

    2004-01-01

    Investigation on ATWS and hypothetical accidents for the Indonesian Multipurpose Reactor RSG-GAS have been undertaken by computer simulation technique. Two computer codes, namely RELAP5 and PARET-ANL, were used as the main tools. The RELAP5 was utilized to perform system analysis while the PARET-ANL code was used to perform the reactor core analysis in more detail. Two different models have been applied as a basis of the simulation: Typical Working Core model (IWC-model) consisting of four regions with different radial power factors; and the hot-channel model consisting of two regions with different radial power factors. Both RELAP5 ad PARET-ANL results showed that in the occurrence of ATWS, failure on fuel element or fuel plate was limited to the region with the most highest power factor. The results also indicated that no high pressure development occurs in that region, so that mechanical damage on the fuel element or other core components due to pressure shock did not happen.(author)

  11. Radioactive particulate release associated with the DOT specification 6M container under hypothetical accident conditions

    International Nuclear Information System (INIS)

    Taylor, J.M.; Raney, P.J.

    1986-02-01

    A testing program was conducted to determine the leakage of depleted uranium dioxide powder (DUO) from the inner containment components of the US Department of Transportation's (DOT) specification 6M container under hypothetical accident conditions. Depleted uranium dioxide was selected as a surrogate for plutonium oxide because of the similarities in the powder characteristics, density and particle size, and because of the special handling and special facilities required for plutonium oxide. The DUO was packaged inside food pack cans in three different configurations inside the 2R vessel of the 6M container. The amount of DUO powder leakage ranged from none detectable ( -7 g) to a high of 1 x 10 -3 g. The combination of gravity, vibration and pressure produced the highest leakage of DUO. Containers that had hermetic seals (leak rates -4 atm cc/min) did not leak any detectable amount ( -7 g) of DUO under the test conditions. Impact forces had no effect on the leakage of particles with the packaging configurations used. 23 refs., 24 figs., 3 tabs

  12. An assessment of the effect of reactor size on hypothetical ore disruptive accidents

    International Nuclear Information System (INIS)

    Buttery, N.E.; Board, S.J.

    1978-01-01

    There is a general tendency towards larger plant sizes, in the interests primarily of economies of scale. In this paper the effect of core size on hypothetical core disruptive accidents (HCDA) is considered, and it is shown that the energy yield increases rapidly with size, primarily due to a tendency towards coherence of voiding in reactors with a large positive void coefficient. Small cores compare favourably in this respect with alternative large designs with low void coefficient cores, because the reduced mass more than compensates for the reduced doppler constant, and they also have a potential advantage in later stages of HCDA (transition phase and after). If energetic HCDA cannot be shown to be unrealistic and if containment of these events is provided as part of the general safety philosophy, then the costs (which may increase disproportionately with yield) of engineering an adequately reliable system needs to be accounted for. Containment costs are only one of many factors which need to be taken into account in optimising the design and so the energy release from a HCDA must take its proper place in the optimisation according to the safety principles and safety case agreed for LMFBRS. (author)

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

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

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

  14. An Analysis of Reactor Structural Response to Fuel Sodium Interaction in a Hypothetical Core Disruptive Accident

    International Nuclear Information System (INIS)

    Suzuki, K.; Tashiro, M.; Sasanuma, K.; Nagashima, K.

    1976-01-01

    This study shows the effect of constraints around FSI zone on FSI phenomena and deformations of reactor structures. SUGAR-PISCES code system has been developed to evaluate the phenomena of FSI and the response of reactor structure. SUGAR calculates the phenomena of FSI. PISCES, developed by Physics International Company in U.S.A, calculates the dynamic response of reactor structure in two-dimensional, time-dependent finite-difference Lagrangian model. The results show that the peak pressure and energy by FSI and the deformation of reactor structures are about twice in case of FSI zone surrounding by blanket than by coolant. The FSI phenomena highly depend on the reactor structure and the realistic configuration around core must be considered for analyzing hypothetical core disruptive accident. In conclusion: FSI phenomena depend highly on constraints around FSI zone, so that the constraints must be dealt with realistically in analytical models. Although a two-dimensional model is superior to a quasi-two-dimensional model. The former needs long calculation time, so it is very expensive using in parametric study. Therefore, it is desirable that the two-dimensional model is used in the final study of reactor design and the quasi-two-dimensional model is used in parametric study. The blanket affects on the acoustic pressure and the deformations of radial structures, but affects scarcely on the upper vessel deformation. The blanket also affects on the mechanical work largely. The core barrel gives scarcely the effects on pressure in single phase but gives highly the effects on pressure in two-phase and deformation of reactor structures in this study. For studying the more realistic phenomena of FSI in the reactor design, the following works should be needed. (i) Spatial Distribution of FSI Region Spatial and time-dependent distribution of fuel temperature and molten fuel fraction must be taken in realistic simulation of accident condition. To this purpose, the code will

  15. Phebus FP. An international severe accident research programme

    International Nuclear Information System (INIS)

    Hardt, P.; Tattegrain, A.

    1995-01-01

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

  16. Research and development strategy on the behavior of containments during severe accidents

    International Nuclear Information System (INIS)

    Lecomte, C.

    1990-06-01

    In case of an hypothetical severe accident leading to core melting, the last barrier preventing radionucleide release in the environnment is the containment of the main reactor building. The French research and development programmes aimed at understanding the containment behavior during severe accidents relate to several domains; some of them are: - assessment of hydrogen behavior - corium behavior and coolability - ultimate resistance of the containments and leaktightness - caracterization of filtered venting procedure. All these aspects are covered by code calculations and experimental developments

  17. 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.)

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  20. Deterministic analyses of severe accident issues

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  1. Jose Cabrera NPP severe accident management activities

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  2. Conclusions on severe accident research priorities

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  3. 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

  4. A review of severe accident assessment

    International Nuclear Information System (INIS)

    Kawashima, Kei

    2000-01-01

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

  5. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

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

  6. Severe accident mitigation through containment design

    International Nuclear Information System (INIS)

    Bergeron, K.D.

    1990-01-01

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

  7. Effects of recent modeling developments in prompt burst hypothetical core disruptive accident calculations

    International Nuclear Information System (INIS)

    Sienicki, J.J.; Abramson, P.B.

    1978-01-01

    The main objective of the development of multifield, multicomponent thermohydrodynamic computer codes is the detailed study of hypothetical core disruptive accidents (HCDAs) in liquid-metal fast breeder reactors. The main contributions such codes are expected to make are the inclusion of detailed modeling of the relative motion of liquid and vapor (slip), the inclusion of modeling of nonequilibrium/nonsaturation thermodynamics, and the use of more detailed neutronics methods. Scoping studies of the importance of including these phenomena performed with the parametric two-field, two-component coupled neutronic/thermodynamic/hydrodynamic code FX2-TWOPOOL indicate for the prompt burst portion of an HCDA that: (1) Vapor-liquid slip plays a relatively insignificant role in establishing energetics, implying that analyses that do not model vapor-liquid slip may be adequate. Furthermore, if conditions of saturation are assumed to be maintained, calculations that do not permit vapor-liquid slip appear to be conservative. (2) The modeling of conduction-limited fuel vaporization and condensation causes the energetics to be highly sensitive to variations in the droplet size (i.e., in the parametric values) for the sizes of interest in HCDA analysis. Care must therefore be exercised in the inclusion of this phenomenon in energetics calculations. (3) Insignificant differences are observed between the use of space-time kinetics (quasi-static diffusion theory) and point kinetics, indicating again that point kinetics is normally adequate for analysis of the prompt burst portion of an HCDA. (4) No significant differences were found to result from assuming that delayed neutron precursors remain stationary where they are created rather than assuming that they move together with fuel. (5) There is no need for implicit coupling between the neutronics and the hydrodynamics/thermodynamics routines, even outside the prompt burst portion

  8. Doppler reactivity uncertainties and their effect upon a hypothetical LOF accident

    International Nuclear Information System (INIS)

    Malloy, D.J.

    1976-01-01

    The statistical uncertainties and the major methodological errors which contribute to the Doppler feedback uncertainty were reviewed and investigated. Improved estimates for the magnitudes of each type of uncertainty were established. The generally applied reactivity feedback methodology has been extended by explicitly treating the coupling effect which exists between the various feedback components. The improved methodology was specifically applied to the coupling of Doppler and sodium void reactivities. In addition, the description of the temperature dependence of the Doppler feedback has been improved by the use of a two-constant formula on a global and regional basis. Feedback and coupling coefficients are presented as a first comparison of the improved and the currently applied methods. Further, the energy release which results from hypothetical disassembly accidents was simulated with a special response surface in the parametric safety evaluation code PARSEC. The impact of the improved feedback methodology and of Doppler coefficient uncertainties was illustrated by the usual parametric relationship between available work-energy and the Doppler coefficient. The work-energy was calculated with the VENUS-II disassembly code and was represented as a response surface in PARSEC. Additionally, the probability distribution for available work-energy, which results from the statistical uncertainty of the Doppler coefficient, was calculated for the current and the improved feedback methodology. The improved feedback description yielded about a 16 percent higher average value for the work-energy. A substantially larger increase is found on the high-yield end of the spectrum: the probability for work-energy above 500 MJ was increased by about a factor of ten

  9. Interaction of radionuclides in severe accident conditions

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  10. United States position on severe accidents

    International Nuclear Information System (INIS)

    Ross, D.F.

    1988-01-01

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

  11. Uncertainties and severe-accident management

    International Nuclear Information System (INIS)

    Kastenberg, W.E.

    1991-01-01

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

  12. The development of severe accident analysis technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  13. A critical review of Jan Beyea's report: A study of some of the consequences of hypothetical reactor accidents at Barsebaeck

    International Nuclear Information System (INIS)

    Gjoerup, H.L.; Hedemann Jensen, P.; Jensen, N.O.; Pejtersen, V.; Lundtang Petersen, E.; Petersen, T.; Thykier-Nielsen, S.; Heikel Vinther, F.

    1978-04-01

    This report contains a critical review of Jan Beyea's report: A study of some of the consequences of hypothetical reactor accidents at Barsebaeck (Princeton University, January 1978). Unreasonable assumptions concerning dry deposition, plume rise, meteorological considerations, dose-response relationship and probability distributions were found in the report. It is found that the conclusions of the Beyea report are the result of a mathematical exercise rather than the results of a realistic risk evaluation for Barsebaeck. (author)

  14. Inventory of programs. Calculation of the isotope inventory after a hypothetical accident at the Cofrentes Nuclear power

    International Nuclear Information System (INIS)

    Albendea, M.

    2014-01-01

    Iberdrola is developing a new application to calculate the inventory of radiological material, then of a hypothetical accident, with the name of inventory. This application allows you to calculate the inventory isotopic, analysers and accurate thermal of all or part of the nucleus of the plant of Cofrentes, even of any single element, based on its history of irradiation and specific periods of decay, since the reactor at any time after the shutdown. (Author)

  15. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  16. 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)

  17. Analysis of hypothetical loss-of-control-arm accidents in HIFAR

    International Nuclear Information System (INIS)

    Connolly, J.W.; Clark, N.

    1986-11-01

    The reactor power transient produced in the HIFAR materials testing reactor upon severance of a central coarse control arm connecting rod and the subsequent pivoting of the arm out of the core has been calculated for a range of reactor conditions likely to be encountered in normal operation. It is concluded that as long as the remaining arms of the control arm bank can be relied on to suppress the post power peak oscillations in power, the reactor will withstand the consequences of such an accident

  18. Point Source contamination approach for hydrological risk assessment of a major hypothetical accident from second research reactor at Inshas site

    International Nuclear Information System (INIS)

    Sadek, M.A.; Tawfik, F.S.

    2002-01-01

    The point source contamination mechanism and the deterministic conservative approach have been implemented to demonstrate the hazards of hydrological pollution due to a major hypothetical accident in the second research reactor at Inshas. The radioactive inventory is assumed to be dissolved in 75% of the cooling water (25% are lost) and comes directly into contact with ground water and moved down gradient. Five radioisotopes(I-129, Sr-90, Ru-106, Cs-134 and Cs-137) of the entire inventory are found to be highly durable and represent vulnerability in the environment. Their downstream spread indices; C max : maximum concentration at the focus of the moving ellipse, delta: pollution duration at different distances, A:polluted area at different distances and X min : safety distance from the reactor, were calculated based on analytical solutions of the convection-dispersion partial differential equation for absorbable and decaying species. The largest downstream contamination range was found for Sr-90 and Ru-106 but still no potential. The geochemical and hydrological parameters of the water bearing formations play a great role in buffering and limiting the radiation effects. These reduce the retention time of the radioisotopes several order of magnitudes in the polluted distances. Sensitivity analysis of the computed pollution ranges shows low sensitivity to possible potential for variations activity of nuclide inventory, dispersivity and saturated thickness and high sensitivity for possible variations in groundwater velocity and retention factors

  19. Severe accident issue resolution -- definition and perspective

    International Nuclear Information System (INIS)

    Harper, F.T.

    1995-01-01

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

  20. Assessment of Radiological and Economic Consequences of a Hypothetical Accident for ETRR-2, Egypt Utilizing COSYMA Code

    International Nuclear Information System (INIS)

    Tawfik, F.S.; Abdel-Aal, M.M.

    2008-01-01

    A comprehensive probabilistic study of an accident consequence assessment (ACA) for loss of coolant accident (LOCA) has accomplished to the second research reactor ETRR-2, located at Inshas Nuclear Research Center, Cairo, Egypt. PC-COSYMA, developed with the support of European Commission, has adopted to assess the radiological and economic consequences of a proposed accident. The consequences of the accident evaluated in case of early and late effects. The effective doses and doses in different organs carried out with and without countermeasures. The force mentioned calculations were required the following studies: the core inventory due to the hypothetical accident, the physical parameters of the source term, the hourly basis meteorological parameters for one complete year, and the population distribution around the plant. The hourly stability conditions and height of atmospheric boundary layers (ABL) of the concerned site were calculated. The results showed that, the nuclides that have short half-lives (few days) give the highest air and ground concentrations after the accident than the others. The area around the reactor requires the early and late countermeasures action after the accident especially in the downwind sectors. Economically, the costs of emergency plan are effectively high in case of applying countermeasures but countermeasures reduce the risk effects

  1. Numerical module for debris behavior under severe accident conditions

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    Science.gov (United States)

    Krauss, U; Mischkowsky, T

    1993-06-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  4. 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)

  5. 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)

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

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  7. 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

  8. Heat transfer phenomena revelant to severe accidents

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. Heat transfer phenomena relevant to severe accidents

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  10. Source terms derived from analyses of hypothetical accidents, 1950-1986

    International Nuclear Information System (INIS)

    Stratton, W.R.

    1987-01-01

    This paper reviews the history of reactor accident source term assumptions. After the Three Mile Island accident, a number of theoretical and experimental studies re-examined possible accident sequences and source terms. Some of these results are summarized in this paper

  11. Thermal hydraulics of CANDU severe accident analysis

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  12. 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

  13. Vaporization of structural materials in severe accidents

    International Nuclear Information System (INIS)

    Lorenz, R.A.

    1982-01-01

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

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

    International Nuclear Information System (INIS)

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

    2012-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2012-06-15

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

  16. 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.

  17. Atmospheric dispersion modeling and radiological safety analysis for a hypothetical accident of Ghana Research Reactor -1 (GHARR-1)

    International Nuclear Information System (INIS)

    Lunguya, J. M.

    2013-06-01

    This work presents the environmental impact analysis of some selected radionuclides released from the Ghana Research Reactor- 1 (GHARR-1) after a hypothetical postulated accidents scenario. The source term was identified and generated from an inventory of radioisotopes released during the accident. Atmospheric transport model was then applied to calculate the total effective dose and how it would be distributed to different organs of the human body as a function of distance downwind. All accident scenarios were selected from GHARR-1 Safety Analysis Report. After the source term was identified the MCNPX code was used to perform the core burnup/depletion analysis. The assumption was made that the activities were released to the atmosphere under a horse design basis accident scenario. The gaussian dose calculation method was applied, coded in Hotspot, a Healthy Physics computer code. This served as the computational tool to perform the atmospheric dispersion modeling and was used to calculate radionuclide concentration at downwind location. Based upon predominant meteorological conditions at the site, the adopted strategy was to use site-specific meteorological data and dispersion modeling to analyze the hypothetical release to the environment of radionuclides and evaluate to what extent such a release may have radiological effects on the public. Final data were processed and presented as Total Effective Dose Equivalent as a function of time and distance of deposition. The results indicate that all the values of Effective dose obtained are far below the regulatory limits, making the use of the reactor safe, even in the case of worst accident scenario where all the fission products were released into the atmosphere. (au)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  19. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  20. Simulation of severe accidents in COTELS experiments

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  1. Overview of severe accident research at KAERI

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  2. 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.

  3. An analysis of severe air transport accidents

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  5. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

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

  7. Studies on the effects of the hypothetic accidents in HTR reactors. (phase 1). HSK 1. Pt. 1

    International Nuclear Information System (INIS)

    Gabriel, H.W.; Redondo, J.A.

    1977-09-01

    The report is an attempt to outline the possibilities of a quantitative, i.e. objective safety assessment. The basic problem here are the controversial opinions an the term of 'risk'. As long as there is no technically and legally acceptable compromise between the opinion 'risk equals extent of damage' and the probabilistic concept, systems in development must follow both approaches. Using the HTR-1160 as an example, studies on the determination of the maximum possible extent of damage in nuclear power plants have been carried out with a special view to determining the rate of the course of extreme accident combinations. Knowledge on this point helps to quantify the chances to prevent the accident itself as well as the chances to protect the population. One basic assumption is that there are no active safety measures to ameliorate the course of the accident (hypothetic accident chains). Consequence analysis is based on nothing but analytically and experimentally validable data, i.e. 'natural law data' on the failure of passive reactor components. Present findings show that HTR reactors can be designed in such manner that accidents with vicinity dose development velocities above 100 rem/5 h are practically impossible. The time history of dose development velocities in the vicinity can be superposed by the course of possible administrative measures. Risk values can then be assessed with sufficient accuracy. (orig./HP) [de

  8. Assessment of uncertainties in severe accident management strategies

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. The retardation effect of structural graphite on the release of fission products in case of hypothetical accidents of HTRs

    International Nuclear Information System (INIS)

    Iniotakis, N.; Decken, C.B. von der

    1982-01-01

    In case of a hypothetical core heat up accident of an HTR the structural graphite of the reactor causes under certain circumstances a very important retardation of the release of fission products into the containment building of the plant. A model is presented which describes the transport phenomena in the graphite structure extensively taking into account specially the macro-structure of the graphite. It is shown by parameter variations under which conditions one can expect a large retardation effect and quantitative values of this retardation, which can be very important, are given. (author)

  10. 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)

  11. In-vessel natural circulation during a hypothetical loss-of-heat-sink accident in the Fast Flux Test Facility

    International Nuclear Information System (INIS)

    Perkins, K.R.; Bari, R.A.; Pratt, W.T.

    1979-05-01

    The capability to remove decay heat from the FFTF core via in-vessel natural circulation has been analyzed for the preboiling phase using a lumped parameter model. The results indicate that boiling will occur in the average fuel assembly for a wide spectrum of initial conditions which appear to be representative of the hypothetical loss-of-heat-sink accident. Two-phase pressure drop calculations indicate that, once the saturation temperature is reached, coolability can only be assured for decay heat levels which are less than 0.5% of the operating power. A review of the limited sodium boiling data indicates that boiling-induced natural circulation may support up to 4% of the operating power, but geometric atypicalities and a large degree of inlet subcooling for the existing data limit the applicability to the loss-of-heat-sink accident in FFTF

  12. 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

  13. 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

  14. Hypothetical accidents at disposal facilities for high-level liquid radioactive wastes and pulps

    International Nuclear Information System (INIS)

    Kabakchi, S.A.; Zagainov, V.A.; Lishnikov, A.A.; Nazin, E.R.

    1994-01-01

    Four accidents are postulated and analyzed for interim storage of high-level, liquid radioactive wastes at a fuel reprocessing facility. Normal waste storage operation is based on wastes stored in steel drums, partially buried in concrete canyons, and equipped with heat exchangers for cooling and ventilation systems for removal of explosive gases and vapors. The accident scenarios analyzed are: (1) shutdown of ventilation with open entrance and exit ventilation pipes, (2) shutdown of ventilation with closed entrance and exit ventilation pipes, (3) shutdown of the cooling system with normally functioning ventilation, and (4) simultaneous cooling and ventilation system failure (worst case). A mathematical model was developed and used to calculate radiation consequences of various accidents. Results are briefly presented for the worst case scenario and compared to an actual accident for model validation. 17 refs., 3 figs., 1 tab

  15. Severe accidents and ESFR design issues

    International Nuclear Information System (INIS)

    Rineiski, A.

    2013-01-01

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

  16. Assessment in marine environment for a hypothetic nuclear accident based on the database of tidal harmonic constants

    International Nuclear Information System (INIS)

    Min, Byung-Il; Periáñez, Raúl; Park, Kihyun; Kim, In-Gyu; Suh, Kyung-Suk

    2014-01-01

    Highlights: • An oceanic dispersion assessment system has been developed. • The developed system is based on a database of tidal harmonic constants. • It used to evaluate pollutant behavior for the hypothetical nuclear accident. • It can predict the pollutant distributions with real-time in the ocean. - Abstract: The eleven nuclear power plants in operation, under construction and a well-planned plant in the east coast of China generally use seawater for reactor cooling. In this study, an oceanic dispersion assessment system based on a database of tidal harmonic constants is developed. This system can calculate the tidal current without a large computational cost, and it is possible to calculate real-time predictions of pollutant dispersions in the ocean. Calculated amplitudes and phases have maximum errors of 10% and 20% with observations, respectively. A number of hypothetical simulations were performed according to varying of the release starting time and duration of pollutant for the six nuclear sites in China. The developed system requires a computational time of one hour for one month of real-time forecasting in Linux OS. Thus, it can use to evaluate rapidly the dispersion characteristics of the pollutants released into the sea from a nuclear accident

  17. CANDU severe accident management guidance update

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  18. CANDU severe accident management guidance update

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-07-01

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

  19. 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.

  20. 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

  1. Method of assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  2. 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.

  3. Severe Accident Test Station Design Document

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  4. Control room habitability during severe accidents

    International Nuclear Information System (INIS)

    Siu, R.P.

    1989-01-01

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

  5. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

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

    1986-08-01

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

  6. Improvement of the severe accident practice tool

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Takahashi, Shunsuke

    2016-01-01

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

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

    International Nuclear Information System (INIS)

    Dennis, A.W.

    1978-01-01

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

  8. Analyses of containment loading by hydrogen burning during hypothetical core meltdown accidents

    International Nuclear Information System (INIS)

    Bracht, K.; Tiltmann, M.

    1983-01-01

    The possibility of occurance of violent hydrogen burning during a LWR meltdown accident and its consequences to containment atmosphere conditions are discussed. Two accident sequences with low and high system pressure during the in-vessel-melt phase of a meltdown accident are considered. In both sequences only deflagration, but no detonation may become possible, presuming homogeneity of the containment atmospheres. In a low pressure szenario the pressure increase due to deflagration will not reach the failure pressure of the containment, if combustion takes place when the flammability limit is reached. For the special situation of a rapid release of steam and hydrogen after a high-pressure failure of a reactor pressure vessel, calculations with a multicompartment code show that the possibility for hydrogen burning does not exist. Thus, an additional augmentation of the steam spike as a consequence of the failure of the pressure vessel cannot occur. (orig.)

  9. 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)

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  11. 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)

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

    International Nuclear Information System (INIS)

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

    2010-01-01

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

  13. 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.

  14. ATHENA2D, Simulation Hypothetical Recriticality Accident in a Thermal Neutron Spectrum

    International Nuclear Information System (INIS)

    1999-01-01

    1 - Description of program or function: ATHENA 2 D was written to simulate a hypothetical water reflood of a highly-damaged light water reactor (such as the Three-Mile-Island Unit-2 after meltdown, with a packed debris bed near the center of the core), but with insufficiently-borated reflood water. A recriticality transient may result because of the potentially more reactive debris bed. ATHENA-2D solves the transient multigroup neutron diffusion equations in (r,z) geometry. Executing in parallel with the transient neutronics, is a single-phase computational fluid dynamics (CFD) model, driven by multichannel thermal hydraulics based on detailed pin models. Numerous PV-Wave procedure files are included on the distribution media, useful for those who already have PV-Wave from Visual Numerics. These procedures are documented in the 'README' files included on the distribution CD. Some reactor lattice computer code such as WIMS-E, CCC-576/WIMSD4, or CCC-656/WIMSD5B is required for the creation of macroscopic cross section libraries, given pin-cell geometries. WIMS-E is a commercial product available from AEA Technologies, England, WIMS is not included on the ATHENA 2 D distribution CD. Several auxiliary routines are included in the package. TFMAX: Utility that searches through ATHENA 2 D binary output to find the maximum fuel temperature over space and time. POST V EL: Utility that searches through ATHENA 2 D binary output to find maximum scalar and flow field values (over space) and outputs normalization factors as a function of time. These results are used to correctly scale animations. CONVT: If executing ATHENA 2 D on a PC under Windows, this utility converts one form of binary output (directly from ATHENA 2 D) to another, which is readable by PV-Wave for Windows (PV-Wave is data animation and visualization software from Visual Numerics, Inc.) CALC M TX: Post-processing utility for calculating the model coefficients for the calculation matrix. 2 - Methods: Both the

  15. 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

  16. Consequences of severe nuclear accidents in Europe

    Science.gov (United States)

    Seibert, Petra; Arnold, Delia; Mraz, Gabriele; Arnold, Nikolaus; Gufler, Klaus; Kromp-Kolb, Helga; Kromp, Wolfgang; Sutter, Philipp

    2013-04-01

    A first part of the presentation is devoted to the consequences of the severe accident in the 1986 Chernobyl NPP. It lead to a substantial radioactive contaminated of large parts of Europe and thus raised the awareness for off-site nuclear accident consequences. Spatial patterns of the (transient) contamination of the air and (persistent) contamination of the ground were studied by both measurements and model simulations. For a variety of reasons, ground contamination measurements have variability at a range of spatial scales. Results will be reviewed and discussed. Model simulations, including inverse modelling, have shown that the standard source term as defined in the ATMES study (1990) needs to be updated. Sensitive measurements of airborne activities still reveal the presence of low levels of airborne radiocaesium over the northern hemisphere which stems from resuspension. Over time scales of months and years, the distribution of radionuclides in the Earth system is constantly changing, for example relocated within plants, between plants and soil, in the soil, and into water bodies. Motivated by the permanent risk of transboundary impacts from potential major nuclear accidents, the multidisciplinary project flexRISK (see http://flexRISK.boku.ac.at) has been carried out from 2009 to 2012 in Austria to quantify such risks and hazards. An overview of methods and results of flexRISK is given as a second part of the presentation. For each of the 228 NPPs, severe accidents were identified together with relevant inventories, release fractions, and release frequencies. Then, Europe-wide dispersion and dose calculations were performed for 2788 cases, using the Lagrangian particle model FLEXPART. Maps of single-case results as well as various aggregated risk parameters were produced. It was found that substantial consequences (intervention measures) are possible for distances up to 500-1000 km, and occur more frequently for a distance range up to 100-300 km, which is in

  17. 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

  18. 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

  19. Sizing of type B package tie-downs on the basis of criteria related to hypothetical road transport accident conditions

    International Nuclear Information System (INIS)

    Phalippou, C.

    1986-01-01

    The aim is to guarantee intactness of the type B package containment system under hypothetical road accident conditions. Some experiments performed in France have led to analytical studies taking into account: a) the head-on collision, which is modelised by a uniform deceleration of 35 g, b) the side-on collision, which is modelised by a colliding object 3 times heavier than the package and an impact at 31.9 km/h. In the first case, the adopted criterion is the holding of the package on the vehicle by the strenght of the stowing members (tie-downs and chocks). In the second case, the adopted criterion is the desired breaking of the tie-downs in order to undamage package containment system; therefore it is assumed that no chock is acting against lateral impacts. Analytical and abacus methods have been developed for sizing the strenght of the stowing members in respect with the two above criteria [fr

  20. Several accidents about ERHRS of CEFR

    International Nuclear Information System (INIS)

    Zhang, D.

    2000-01-01

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

  1. Monitoring and operation system for severe accidents

    International Nuclear Information System (INIS)

    Fukui, Toshiki; Niida, Shinji; Kato, Yumeto

    2017-01-01

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

  2. Numerical Analysis of Molten Corium Dispersion during Hypothetical High-Pressure Accidents in APR1400 Nuclear Power Plant

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Ha, Kwang Soon; Kim, Sang Baik; Kim, Hee Dong; Jeong, Jae Sik

    2010-01-01

    During a hypothetical high-pressure accident in a nuclear power plant (NPP), molten corium can be ejected through a breach of a reactor pressure vessel (RPV) and dispersed by the following jet of a high pressure steam in the RPV. The dispersed corium is fragmented into smaller droplets in a reactor cavity of the NPP by the steam jet with very high velocity and is released into the upper compartment of the NPP by an overpressure in the cavity. The heat-carrying fragments of the corium transfer the thermal energy to the ambient air in the containment and react chemically with steam and generate hydrogen which may be burnt in the containment. The thermal loads from the ejected molten corium on the containment which is called direct containment heating (DCH) can threaten the integrity of the containment. New generation NPPs such as APR1400 and EPR have been designed in consideration of reducing the possibility of the containment failure from the DCH. In order for that, APR1400 has a convolute-type corium chamber connected to the reactor cavity. In the case of EPR, severe-accident dedicated depressurization valves are installed to preclude a high pressure melt ejection (HPME). DCH in a NPP containment is related to many physical phenomena such as multi-phase hydrodynamics, thermodynamics and chemical reaction. In the evaluation of the DCH load, the melt dispersion rates depending on the RPV pressure are the most important parameter. Mostly, DCH was evaluated by using lumped-analysis codes with some correlations obtained from experiments for the dispersion rates. The corium dispersion rates for many types of the NPP containments had been obtained by experiments in 90s. And some correlations from the experimental data were developed. As mentioned above, APR1400 has a corium chamber to reduce the corium dispersion rate. But there is no experimental data for the dispersion rate specific to the APR1400 cavity geometry. So its performance for capturing of the dispersed corium

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

    International Nuclear Information System (INIS)

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

    1993-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-12-15

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

  5. 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)

  6. 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)

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

    International Nuclear Information System (INIS)

    Abe, Kiyoharu.

    1995-05-01

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

  8. ACR-1000 design provisions for severe accidents

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  9. 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

  10. Licensing aspects in the verification of the SNR 300 design concept against hypothetical accidents

    International Nuclear Information System (INIS)

    Kugler, E.; Wiesner, S.

    1976-01-01

    The German prototype of a fast breeder reactor, the SNR 300, is being built near Kalkar on the Lower Rhine. It is a loop-type fast sodium-cooled reactor, designed and constructed by Interatom, Bensberg. Experiences gained from the first phase of construction are described. The report is restricted to the aspects of the SNR 300 design against a core disruptive accident (CDA) and its consequences and to the difficulties having arisen in the verification of the design concept so far. Some examples of the detailed design are described and discussed from the licensing authority's point of view showing that the difficulties have been typical for a prototype reactor subjected to a regular licensing procedure

  11. Big Rock Point severe accident management strategies

    International Nuclear Information System (INIS)

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

    1996-01-01

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

  12. 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

  13. Improvement of severe accident analysis method for KSNP

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-03-15

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

  14. SARNET: Severe accident research network of excellence

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Transient debris freezing and potential wall melting during a severe reactivity initiated accident experiment

    International Nuclear Information System (INIS)

    El-Genk, M.S.; Moore, R.L.

    1981-01-01

    It is important to light water reactor (LWR) safety analysis to understand the transient freezing of molten core debris on cold structures following a hypothetical core meltdown accident. The purpose of this paper is to (a) present the results of a severe reactivity initiated accident (RIA) in-pile experiment with regard to molten debris distribution and freezing following test fuel rod failure, (b) analyze the transient freezing of molten debris (primarily a mixture of UO/sub 2/ fuel and Zircaloy cladding) deposited on the inner surface of the test shroud wall upon rod failure, and (c) assess the potential for wall melting upon being contacted by the molten debris. 26 refs

  16. Severe accident considerations for modern KWU-PWR plants

    International Nuclear Information System (INIS)

    Eyink, J.

    1987-01-01

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

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  18. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

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

    2000-01-01

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

  19. Development of the simulation system IMPACT for analysis of nuclear power plant severe accidents

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-01-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system IMPACT for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT's distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data

  20. 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

  1. COSYMA, a mainframe and PC program package for assessing the consequences of hypothetical accidents

    International Nuclear Information System (INIS)

    Jones, J.A.; Hasemann, I.; Steen, J. van der

    1996-01-01

    COSYMA (Code System from MARIA) is a program package for assessing the off-site consequences of accidental releases of radioactive material to atmosphere, developed as part of the European Commission's MARIA programme (Methods for Assessing the Radiological Impact of Accidents). COSYMA represents a fusion of ideas and modules from the Forschungszetrum Karlsruhe program system UFOMOD, the National Radiological Protection Board program MARC and new model developments together with data libraries from other MARIA contractors. Mainframe and PC versions of COSYMA are distributed to interested users by arrangement with the European Commission. The system was first released in 1990, and has subsequently been updated. The program system uses independent modules for the different parts of the analysis, and so permits a flexible problem-oriented application to different sites, source terms, emergency plans and the needs of users in the various parts of Europe. Users of the mainframe system can choose the most appropriate combination of modules for their particular application. The PC version includes a user interface which selects the required modules for the endpoints specified by the user. This paper describes the structure of the mainframe and PC versions of COSYMA, and summarises the models included in them. The mainframe or PC versions of COSYMA have been distributed to more than 100 organisations both inside and outside the European Union, and have been used in a wide variety of applications. These range from full PRA level 3 analyses of nuclear power and research reactors to investigations on advanced containment concepts and the preplanning of off-site emergency actions. Some of the experiences from these applications are described in the paper. An international COSYMA user group has been established to stimulate communication between the owners, developers and users of the code and to serve as a reference point for questions relating to the code. The group produces

  2. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

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

    2006-01-01

    Full text: Atomic Energy of Canada Limited (AECL) developed the Advanced CANDU Reactor-TM700 (ACR-700TM) as an evolutionary advancement of the current CANDU 6R reactor. As further advancement of the ACR design, AECL is currently developing the ACR-1000TM for the Canadian and international market. The ACR-1000 is aimed at producing electrical power for a capital cost and a unit-energy cost significantly less than that of the current generation of operating nuclear plants, while achieving shorter construction schedule, high plant capacity factor, improved operations and maintenance, increased operating life. and enhanced safety features. The reference ACR-1000 plant design is based on an integrated two-unit plant, using enriched fuel and light-water coolant, with each unit having a nominal gross output of about 1200 MWe. This paper presents the ACR-1000 features that are designed to mitigate limited core damage and severe core damage states, including core retention within vessel, core damage termination, and containment integrity maintenance. Core retention within vessel in CANDU-type reactors includes both retention within fuel channels, and retention within the calandria vessel. The moderator heavy water in the ACR-1000 calandria vessel, as in any other CANDU-type reactor, provides ample heat removal capacity in severe accidents. The ACR-1000 calandria vessel design permits for passive rejection of decay heat from the moderator to the shield water. Also, the calandria vessel will be designed for debris retention. Core damage termination is achieved by flooding of the core components with water and keeping them flooded thereafter. Successful termination can be achieved in the fuel channels, calandria vessel or calandria vault by water supply by the Long Term Cooling (LTC) pumps and by gravity feed from the Reserve Water System. The ACR-1000 containment is required to withstand external events such as earthquakes, tornados, floods and aircraft crashes. Containment

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

    International Nuclear Information System (INIS)

    Carbajo, J.J.

    1999-01-01

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

  4. 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)

  5. The numerical simulation of the WWER-440/V-213 reactor pressure vessel internals response to maximum hypothetical large break loss of coolant accident

    International Nuclear Information System (INIS)

    Hermansky, P.; Krajcovic, M.

    2012-01-01

    The reactor internals are designed to ensure cooling of the fuel, to ensure the movement of emergency control assemblies under all operating conditions including accidents and facilitate removal of the fuel and of the internals following an accident This paper presents results of the numerical simulation of the WWER-440/V213 reactor vessel internals dynamic response to maximum hypothetical Large-Break Loss of Coolant Accident. The purpose of this analysis is to determine the reactor vessel internals response due to rapid depressurization and to prove no such deformations occur in the reactor vessel internals which would prevent timely and proper activation of the emergency control assemblies. (Authors)

  6. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

  9. Large scale experiments simulating hydrogen distribution in a spent fuel pool building during a hypothetical fuel uncovery accident scenario

    Energy Technology Data Exchange (ETDEWEB)

    Mignot, Guillaume; Paranjape, Sidharth; Paladino, Domenico; Jaeckel, Bernd; Rydl, Adolf [Paul Scherrer Institute, Villigen (Switzerland)

    2016-08-15

    Following the Fukushima accident and its extended station blackout, attention was brought to the importance of the spent fuel pools' (SFPs) behavior in case of a prolonged loss of the cooling system. Since then, many analytical works have been performed to estimate the timing of hypothetical fuel uncovery for various SFP types. Experimentally, however, little was done to investigate issues related to the formation of a flammable gas mixture, distribution, and stratification in the SFP building itself and to some extent assess the capability for the code to correctly predict it. This paper presents the main outcomes of the Experiments on Spent Fuel Pool (ESFP) project carried out under the auspices of Swissnuclear (Framework 2012–2013) in the PANDA facility at the Paul Scherrer Institut in Switzerland. It consists of an experimental investigation focused on hydrogen concentration build-up into a SFP building during a predefined scaled scenario for different venting positions. Tests follow a two-phase scenario. Initially steam is released to mimic the boiling of the pool followed by a helium/steam mixture release to simulate the deterioration of the oxidizing spent fuel. Results shows that while the SFP building would mainly be inerted by the presence of a high concentration of steam, the volume located below the level of the pool in adjacent rooms would maintain a high air content. The interface of the two-gas mixture presents the highest risk of flammability. Additionally, it was observed that the gas mixture could become stagnant leading locally to high hydrogen concentration while steam condenses. Overall, the experiments provide relevant information for the potentially hazardous gas distribution formed in the SFP building and hints on accident management and on eventual retrofitting measures to be implemented in the SFP building.

  10. On severe accident hydrogen behaviour in Loviisa

    International Nuclear Information System (INIS)

    Okkonen, T.

    1996-02-01

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

  11. On severe accident hydrogen behaviour in Loviisa

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-02-01

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

  12. 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)

  13. Design and Development of a Severe Accident Training System

    International Nuclear Information System (INIS)

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

    2005-01-01

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

  14. 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)

  15. PBDOWN - a computer code for simulating core material discharge and thermal to mechanical energy conversion in LMFBR hypothetical accidents

    International Nuclear Information System (INIS)

    Royl, P.

    1981-01-01

    PBDOWN is a computer code that simulates the blowdown of confined boiling materials ('pools') into a colder upper coolant plenum as time dependent ejection and expansion with consideration of a few selected exchange processes. Its application is restricted to situations resulting from hypothetical loss of flow (LOF) accidents in LMFBR's, where enough voiding has occured, that in core sodium vapor pressures become negligible. PBDOWN considers one working fluid for the discharge process (either fuel or steel) and a maximum of two working fluids (either fuel and sodium or steel and sodium) for the expansion process in the upper coolant plenum. Entrainment of sodium at the accelerated bubble liquid interfaces is mechanistically calculated by a Taylor instability entrainment model. Simulation of a hemispherical expansion form together with this mechanistic entrainment model gives a new integrated calculation of the time dependent sodium mass in the bubble. The paper summarizes the basic equations and assumptions of this computer model. Sample results compare different heat transfer and Na entrainment models during steel and fuel driven discharge processes. Mechanistic sodium entrainment simulation for SNR-type reactors coupled with a realistic heat transfer model is shown to reduce the integral mechanical work potential by a factor of 1.3 to 2.0 over the isentropic energy of the discharge working fluids. (orig.)

  16. Experimental analysis of the behaviour of iodine in the event of hypothetical accidents. Final report. Pt. 1

    International Nuclear Information System (INIS)

    Richter, F.; Rippel, R.; Proebstle, G.; Fernholz, O.

    1986-01-01

    Experiments have been performed simulating hypothetical core-melt accidents in order to determine droplet-bound transport of radio-nuclides. Different measurement methods have been applied to evaluate steam moisture and droplet size distribution, the carry-over factor of a tracer substance, and, to some extent, droplet velocity, under atmospheric sump water boiling conditions. Part flow analysis yields carry-over factor values on the order of magnitude 10 -5 . Thus it is smaller than would be expected from visual measurements of steam moisture in the main flow, a result which is due to droplet velocity characteristics which limit the carry-over through openings. Results distinctly show that steam moisture (10 -3 up to 7x10 -5 , depending on the distance from the sump) and the droplet size (4-57 μm) can only be used as a source term. In order to evaluate the quantity released from a leakage, a supplementary investigation of droplet carry-over mechanisms will be required. (orig.) [de

  17. Assessment of radiological impact due to a hypothetical core disruptive accident for PFBR using an advanced atmospheric dispersion system

    International Nuclear Information System (INIS)

    Srinivas, C.V.; Venkatesan, R.; Natarajan, A.

    2004-01-01

    Radiological impact due to air borne effluent dispersion from a hypothetical Core Disruptive Accident (CDA) scenario for Prototype Fast Breeder Reactor (PFBR) at Kalpakkam coastal site is estimated using an advanced system consisting of a 3-d meso-scale atmospheric model and a random walk particle dispersion model. A simulation of dispersion for CDA carried out for a typical summer day on 24th May 2003 predicted development of land-sea breeze circulation and Thermal Internal Boundary Layer (TIBL) at Kalpakkam site, which have been confirmed by observations. Analysis of dose distribution corresponding to predicted atmospheric conditions shows maximum dose from stack releases beyond the site boundary at about 4 km during TIBL fumigation and stable conditions respectively. A multi mode spatial concentration distribution has been noticed with diurnal meandering of wind under land sea breeze circulation. Over a meso-scale range of 25 km, turning of plume under sea breeze and maximum concentration along plume centerline at distances of 3 to 10 km have been noticed. The study has enabled to simulate the more complex meteorological situation that is actually present at the site. (author)

  18. Hypothetical core disruptive accident analysis of a 2000 MWsub(e) liquid metal cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Struwe, D.

    1977-12-01

    A structural phase diagram for hypothetical core disruptive accidents (HCDA) has been developed based on a variety of analyses for different LMFBR's. The intention was to identify the strategic phases of HCDA's important with regard to safety aspects of the plant. These phases are investigated in detail for a 2,000 MWsub(e) LMFBR (SNR-2,000). Characteristic data of SNR-2,000 are discussed concerning their influence on safety analysis. Reasons for the choice of model parameters for special phenomena as fuel coolant interaction, fuel pin failure mechanisms and sodium voiding are given. The results of calculations with CAPRI-2, HOPE and KADIS are analyzed for possibilities to enter energetic core disassembly with consequences, making power values below 2,000 MWsub(e) necessary. Investigation of these results shows that the expected consequences do not lead to design requirements, restricting the magnitude of the electrical power output of LMFBR's to values below 2,000 MWsub(e). Therefore, consequences of HCDA's are principal not expected to limit the feasibility of conventional core design of this order of magnitude. (orig.) [de

  19. Dose calculations for severe LWR accident scenarios

    International Nuclear Information System (INIS)

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

    1984-05-01

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

  20. Calculation of spent fuel pool severe accident with MELCOR

    International Nuclear Information System (INIS)

    Deng Jian; Xiang Qing'an; Zhou Kefeng

    2014-01-01

    A calculation model was established for spent fuel pool (SFP) using MELCOR code to study the severe accident phenomena caused by the long term station black-out (SBO), including spent fuel heatup, zirconium cladding oxidation, and the injection into SFP to mitigate the severe accident. The results show that the severe accident progression is slow and relates directly with the initial water level in SFP. It is illustrated that the injection into SFP is one of the best mitigated measures for the SFP severe accident. (authors)

  1. 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

  2. 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)

  3. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

  5. 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)

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  8. Strategies for the prevention and mitigation of severe accidents

    International Nuclear Information System (INIS)

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

    1999-01-01

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

  9. Probability estimation of potential harm to human health and life caused by a hypothetical nuclear accident at the nuclear power plant

    International Nuclear Information System (INIS)

    Soloviov, Vladyslav; Pysmenniy, Yevgen

    2015-01-01

    This paper describes some general methodological aspects of the assessment of the damage to human life and health caused by a hypothetical nuclear accident at the nuclear power plant (NPP). Probability estimation of death (due to cancer and non-cancer effects of radiation injury), disability and incapacity of individuals were made by taking into account the regulations of Ukraine. According to the assessment, the probability of death due to cancer and non-cancer effects of radiation damage to individuals who received radiation dose of 1 Sv is equal to 0.09. Probability of disability of 1, 2 or 3 group regardless of the radiation dose is 0.009, 0.0054, 0.027, respectively. Probability of temporary disability of the individual who received dose equal to 33 mSv (the level of potential exposure in a hypothetical nuclear accident at the NPP) is equal 0.16. This probability estimation of potential harm to human health and life caused by a hypothetical nuclear accident can be used for NPP in different countries using requirements of regulations in these countries. And also to estimate the amount of insurance payments due to the nuclear damage in the event of a nuclear accident at the NPP or other nuclear industry enterprise. (author)

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

    OpenAIRE

    SONG, JIN HO; KIM, TAE WOON

    2014-01-01

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

  11. 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)

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

    International Nuclear Information System (INIS)

    Frid, W.; Sandervaag, O.

    1997-01-01

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

  13. Use of probabilistic safety analyses in severe accident management

    International Nuclear Information System (INIS)

    Neogy, P.; Lehner, J.

    1991-01-01

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

  14. France-Japan collaboration on the severe accident studies for ASTRID. Outcomes and future work program

    International Nuclear Information System (INIS)

    Serre, F.; Bertrand, F.; Bachrata, A.; Marie, N.; Kubo, Shigenobu; Kamiyama, Kenji; Carluec, B.; Farges, B.; Koyama, K.

    2017-01-01

    The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator of GenIV sodium-cooled fast reactor (SFR) designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are first to prevent the core melting, in particular by the development of an innovative core (named CFV core) with low void worth and complementary safety prevention devices, and second, to enhance the reactor resistance to severe accidents by design. In order to mitigate the consequences of hypothetical core melting situations, specific provisions (mitigation devices) are added to the core and to the reactor. To meet these ASTRID objectives, a large R and D program was launched in the Severe Accident domain by the CEA, with collaboration of AREVA NP, JAEA, MFBR and MHI organizations, in the frame of the France-Japan ASTRID and SFRs collaboration agreement. This R and D program covers exchanges on severe accident conditions to be studied for the SFR safety cases, the methodology to study these situations, ASTRID severe accident simulations, the comparison and understanding of the ASTRID and JSFR reactor behavior under these situations, the development and adaptation of simulation tools, and, despite an already large existing experimental database, a complementary experimental program to improve the knowledge and reduce the uncertainties. This paper will present the collaboration work performed on the Severe Accidents studies. (author)

  15. Atmospheric transport of radioactive debris to Norway in case of a hypothetical accident related to the recovery of the Russian submarine K-27

    International Nuclear Information System (INIS)

    Bartnicki, Jerzy; Amundsen, Ingar; Brown, Justin; Hosseini, Ali; Hov, Øystein; Haakenstad, Hilde; Klein, Heiko; Lind, Ole Christian; Salbu, Brit; Szacinski Wendel, Cato C.; Ytre-Eide, Martin Album

    2016-01-01

    The Russian nuclear submarine K-27 suffered a loss of coolant accident in 1968 and with nuclear fuel in both reactors it was scuttled in 1981 in the outer part of Stepovogo Bay located on the eastern coast of Novaya Zemlya. The inventory of spent nuclear fuel on board the submarine is of concern because it represents a potential source of radioactive contamination of the Kara Sea and a criticality accident with potential for long-range atmospheric transport of radioactive particles cannot be ruled out. To address these concerns and to provide a better basis for evaluating possible radiological impacts of potential releases in case a salvage operation is initiated, we assessed the atmospheric transport of radionuclides and deposition in Norway from a hypothetical criticality accident on board the K-27. To achieve this, a long term (33 years) meteorological database has been prepared and used for selection of the worst case meteorological scenarios for each of three selected locations of the potential accident. Next, the dispersion model SNAP was run with the source term for the worst-case accident scenario and selected meteorological scenarios. The results showed predictions to be very sensitive to the estimation of the source term for the worst-case accident and especially to the sizes and densities of released radioactive particles. The results indicated that a large area of Norway could be affected, but that the deposition in Northern Norway would be considerably higher than in other areas of the country. The simulations showed that deposition from the worst-case scenario of a hypothetical K-27 accident would be at least two orders of magnitude lower than the deposition observed in Norway following the Chernobyl accident. - Highlights: • Long-term meteorological database has been developed for atmospheric dispersion. • Using this database, the worst case meteorological scenarios have been selected. • Mainly northern parts of Norwegian territory will be

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

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    Purpose Bus safety is a concern not only in developing countries, but also in the U.S. and Europe. In Denmark, disentangling risk factors that are positively or negatively related to bus accident severity and injury occurrence to bus passengers can contribute to promote safety as an essential...... principle of sustainable transit and advance the vision “every accident is one too many”. Methods Bus accident data were retrieved from the national accident database for the period 2002–2011. A generalized ordered logit model allows analyzing bus accident severity and a logistic regression enables...... examining occurrence of injury to bus passengers. Results Bus accident severity is positively related to (i) the involvement of vulnerable road users, (ii) high speed limits, (iii) night hours, (iv) elderly drivers of the third party involved, and (v) bus drivers and other drivers crossing in yellow or red...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-07-01

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

  18. Containment loading during severe core damage accidents

    International Nuclear Information System (INIS)

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

    1984-11-01

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

  19. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    Energy Technology Data Exchange (ETDEWEB)

    Tso, C.F. [Arup (United Kingdom); Hueggenberg, R. [Gesellschaft fuer Nuklear-Behaelter mbH (Germany)

    2004-07-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work.

  20. Evaluation of finite element codes for demonstrating the performance of radioactive material packages in hypothetical accident drop scenarios

    International Nuclear Information System (INIS)

    Tso, C.F.; Hueggenberg, R.

    2004-01-01

    Drop testing and analysis are the two methods for demonstrating the performance of packages in hypothetical drop accident scenarios. The exact purpose of the tests and the analyses, and the relative prominence of the two in the license application, may depend on the Competent Authority and will vary between countries. The Finite Element Method (FEM) is a powerful analysis tool. A reliable finite element (FE) code when used correctly and appropriately, will allow a package's behaviour to be simulated reliably. With improvements in computing power, and in sophistication and reliability of FE codes, it is likely that FEM calculations will increasingly be used as evidence of drop test performance when seeking Competent Authority approval. What is lacking at the moment, however, is a standardised method of assessing a FE code in order to determine whether it is sufficiently reliable or pessimistic. To this end, the project Evaluation of Codes for Analysing the Drop Test Performance of Radioactive Material Transport Containers, funded by the European Commission Directorate-General XVII (now Directorate-General for Energy and Transport) and jointly performed by Arup and Gesellschaft fuer Nuklear-Behaelter mbH, was carried out in 1998. The work consisted of three components: Survey of existing finite element software, with a view to finding codes that may be capable of analysing drop test performance of radioactive material packages, and to produce an inventory of them. Develop a set of benchmark problems to evaluate software used for analysing the drop test performance of packages. Evaluate the finite element codes by testing them against the benchmarks This paper presents a summary of this work

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

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  2. An overview of selected severe accident research and applications

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  3. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    International Nuclear Information System (INIS)

    1989-09-01

    This Standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The specific radionuclides considered in the Standard are those associated with substances having the greatest potential for becoming airborne in reactor accidents (eg, tritium (HTO), noble gases and their daughters (Kr-Rb, Xe-Cs), and radioiodines (I)); and certain radioactive particulates (eg, Cs, Ru, Sr, Te) that may become airborne under exceptional circumstances

  4. Analyses of severe accident scenarios in RBMK-1500

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    2013-12-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-01-01

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

  7. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.

    1988-07-01

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

  8. 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)

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

    International Nuclear Information System (INIS)

    Golshan, Mina

    2013-01-01

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

  10. Aspects of severe accidents in transmutation systems

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  11. 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

  12. Thermal-hydraulic uncertainties affecting severe accident progression

    International Nuclear Information System (INIS)

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

    1984-01-01

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

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

    International Nuclear Information System (INIS)

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

    1989-01-01

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

  14. Addressing severe accidents in the CANDU 9 design

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  15. 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)

  16. Airborne concentrations of radioactive materials in severe accidents

    International Nuclear Information System (INIS)

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

    1989-01-01

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

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

    International Nuclear Information System (INIS)

    Hodge, S.A.

    1983-01-01

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

  18. 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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  20. 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)

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

    International Nuclear Information System (INIS)

    Sadek, Sinisa; Amizic, Milan; Grgic, Davor

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

    2013-01-01

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

  3. Research on sever accident emergency simulation system for CPR1000

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    Science.gov (United States)

    Iyanda, Ayodeji E

    2018-05-28

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

  5. 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

  6. Comparative analysis of a hypothetical 0.1 $/SEC transient overpower accident in an irradiated LMFBR core using different computer models

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Fremont, R. de; Renard, A.

    1982-01-01

    The Report gives the results of comparative calculations performed by the Whole Core Accident Codes Group which is a subgroup of the Safety Working Group of the Fast Reactor Coordinating Committee for a hypothetical transient overpower accident in an irradiated LMFBR core. Different computer codes from members of the European Community and the United States were used. The calculations are based on a Benchmark problem, using commonly agreed input data for the most important phenomena, such as the fuel pin failure threshold, FCl parameters, etc. Beside this, results with alternative assumptions for theoretical modelling are presented with the scope to show in a parametric way the influence of more advanced modelling capabilities and/or better (so-called best estimate) input data for the most important phenomena on the accident sequences

  7. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

    Nielsen, F.; Thykier-Nielsn, S.

    1987-03-01

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

  8. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  9. Trans-oceanic transport of {sup 137}Cs from the Fukushima nuclear accident and impact of hypothetical Fukushima-like events of future nuclear plants in Southern China

    Energy Technology Data Exchange (ETDEWEB)

    Wai, Ka-Ming, E-mail: bhkmwai@cityu.edu.hk [Department of Geological and Mining Engineering and Sciences, Michigan Technological University, Houghton, MI (United States); Department of Physics and Material Science, City University of Hong Kong, Hong Kong (China); Yu, Peter K.N. [Department of Physics and Material Science, City University of Hong Kong, Hong Kong (China)

    2015-03-01

    A Lagrangian model was adopted to assess the potential impact of {sup 137}Cs released from hypothetical Fukushima-like accidents occurring on three potential nuclear power plant sites in Southern China in the near future (planned within 10 years) in four different seasons. The maximum surface (0–500 m) {sup 137}Cs air concentrations would be reached 10 Bq m{sup −3} near the source, comparable to the Fukushima case. In January, Southeast Asian countries would be mostly affected by the radioactive plume due to the effects of winter monsoon. In April, the impact would be mainly on Southern and Northern China. Debris of radioactive plume (∼ 1 mBq m{sup −3}) would carry out long-range transport to North America. The area of influence would be the smallest in July due to the frequent and intense wet removal events by trough of low pressure and tropical cyclone. The maximum worst-case areas of influence were 2382000, 2327000, 517000 and 1395000 km{sup 2} in January, April, July and October, respectively. Prior to the above calculations, the model was employed to simulate the trans-oceanic transport of {sup 137}Cs from the Fukushima nuclear accident. Observed and modeled {sup 137}Cs concentrations were comparable. Sensitivity runs were performed to optimize the wet scavenging parameterization. The adoption of higher-resolution (1° × 1°) meteorological fields improved the prediction. The computed large-scale plume transport pattern over the Pacific Ocean was compared with that reported in the literature. - Highlights: • A Lagrangian model was used to predict the dispersion of {sup 137}Cs from plant accident. • Observed and modeled {sup 137}Cs concentrations were comparable for the Fukushima accident. • The maximum surface concentrations could reach 10 Bq m{sup −3} for the hypothetical case. • The hypothetical radiative plumes could impact E/SE Asia and N. America.

  10. Case study of the effects of hypothetical nuclear power plant accident to the northern food chain of lichen-reindeer-man

    Energy Technology Data Exchange (ETDEWEB)

    Leppaenen, A.P.; Solatie, D. [Radiation and Nuclear Safety Authority - STUK (Finland); Paatero, J. [Finnish Meteorological Institute (Finland)

    2014-07-01

    There are plans to open a new nuclear power plant in Northern Finland at Pyhaejoki. The currently planned reactor type is AES 2006 built by Rosenergoatom. The power output of the AES 2006 is 1200 MWe. In a hypothetical reactor accident at Pyhaejoki large amounts of radioactivity would be released to the environment in Northern Europe. With suitable wind conditions the contaminants would contaminate large areas in the Euro-Arctic region in Northern Scandinavia and in Kola Peninsula. Northern parts of Scandinavia belongs to the sub-arctic region where reindeer herding is an important livelihood for the local and for the indigenous Sami people. As a results of the CEEPRA-project ('Collaboration Network on Environmental Radiation Protection and Research') funded by the EU's Kolarctic ENPI CBC program estimated a possible fallout to Finnish Lapland from a hypothetical nuclear power plant accident occurring at the planned site. Lichen-reindeer-man food chain is an important food chain to the people living in Lapland from traditional and from economical point of views. The food chain is known to enrich radioactive contaminants efficiently. In case of nuclear fallout this food chain would be one of the primary sources of {sup 137}Cs into the inhabitants in Northern regions. The food chain has been well-studied where studies began in the 1960's and was intensified after the Chernobyl accident. This study concentrates on the effects caused by the hypothetical accident, occurring at the planned Pyhaejoki power plant, to the lichen-reindeer-man food chain. The transfer of {sup 137}Cs and {sup 134}Cs to the reindeer meat and possible doses to the man will be estimated. Document available in abstract form only. (authors)

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

    International Nuclear Information System (INIS)

    Koenig, L.N.

    1986-01-01

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

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

    International Nuclear Information System (INIS)

    Su Changsong

    2010-01-01

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

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

    African Journals Online (AJOL)

    ADOWIE PERE

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

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  15. 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).

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  17. Evaluation of strategies for severe accident prevention and mitigation

    International Nuclear Information System (INIS)

    Tokarz, R.

    1989-01-01

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

  18. 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

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

    International Nuclear Information System (INIS)

    Fernandes, A.; Nakata, H.

    1991-01-01

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

  20. MELCOR DB Construction for the Severe Accident Analysis DB

    International Nuclear Information System (INIS)

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

    2011-01-01

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

  1. 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

  2. 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

  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. Statistical modelling of the frequency and severity of road accidents

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

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

  6. 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

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

    International Nuclear Information System (INIS)

    Gustavsson, Veine

    2002-11-01

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

  8. 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

  9. 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

  10. Instrumentation for the follow-up of severe accidents

    International Nuclear Information System (INIS)

    Munoz Sanchez, A.; Nino Perote, R.

    2000-01-01

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

  11. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehgal, B.R.

    2001-01-01

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

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

    International Nuclear Information System (INIS)

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

    2017-01-01

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

  13. 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

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

    International Nuclear Information System (INIS)

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

    2006-01-01

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

  15. OSSA. A second generation of severe accident management

    International Nuclear Information System (INIS)

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

    2009-01-01

    Nowadays the severe accident and their management are an integrated part of the new generation of power plants. The EPR, as the third generation of nuclear plants, includes both systems and instrumentation to mitigate a severe accident, but also a new generation of severe accident management guidelines: the OSSA. Severe accident management guidelines are highly dependent on human means available: emergency organization actors, training and knowledge shall be taken in consideration in an innovative way. Their impacts on ergonomy and content of the document lead to a new generation of guidelines with several innovative features. This second generation of severe accident management guidelines was developed in parallel with the PSA level 2, the human reliability analyses, the validation and verification process, the severe accident simulator progresses. By taking in consideration this variety of input the OSSA were developed in a user aspect orientation. For example in the OSSA a larger responsibility is given to the operational crew to better support the technical support group evaluation. Their existing knowledge of the plant and of the systems and instrumentation is used. This collaboration work implies a strong communication tool that has been developed to enhance the permanent communication within the emergency organization, but although to ensure the main up-to-date information for evaluation will be available where required. The entry condition is based on a strong and stand alone diagnostic for all plant states, that uses in particular a curve of core exit temperature as a function of primary pressure for a fixed core cladding temperature, or its equivalent in term of containment conditions. It ensures relatively consistent core conditions on entry. A first criterion for ultimate final primary depressurization is provided, ensuring all attempts to reflood the core with the available means have been ensured before the OSSA entry condition is reached. This

  16. Severe accidents and terrorist threats at nuclear reactors

    International Nuclear Information System (INIS)

    Pollack, G.L.

    1987-01-01

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

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

    Science.gov (United States)

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

    2014-02-01

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

  18. 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

  19. Drug use and the severity of a traffic accident

    NARCIS (Netherlands)

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

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

  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. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  2. Nuclear Reactor RA Safety Report, Vol. 16, Maximum hypothetical accident; Izvestaj o sigurnosti nuklearnog reaktora RA, Knjiga 16, Maksimalni hipoteticki akcident

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1986-11-15

    Fault tree analysis of the maximum hypothetical accident covers the basic elements: accident initiation, phase development phases - scheme of possible accident flow. Cause of the accident initiation is the break of primary cooling pipe, heavy water system. Loss of primary coolant causes loss of pressure in the primary circuit at the coolant input in the reactor vessel. This initiates safety protection system which should automatically shutdown the reactor. Separate chapters are devoted to: after-heat removal, coolant and moderator loss; accident effects on the reactor core, effects in the reactor building, and release of radioactive wastes. [Serbo-Croat] Sema granjanja za maksimalni hipoteticki akcident obuhvata osnovne elemente: pocetak akcidenta, faze razvoja akcidenta i stablo razvoja - sema potencijalnih akcidentnih tokova. Uzrok pocetka akcidenta je pucanje cevovoda primarnog rashladnog sistema jezgra, sistema teske vode. Gubitak primarnog hladioca izaziva pad pritiska u primarnom sistemu hladjenja na ulazu u reaktorski sud. Ovaj poremecaj pobudjuje sigurnosno kolo zastite koje automatski treba da prekine rad reaktora. Posebno je razmatrano generisanje zaostale snage, isticanje hladioca i moderatora, efekti akcidenta na jezgro, efekti u zgradi reaktora, oslobadjanje radioaktivnih produkata.

  3. Source term analyses under severe accidents for KNGR

    Energy Technology Data Exchange (ETDEWEB)

    Song, Yong Mann; Park, Soo Yong

    2001-03-01

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

  4. Development of SPEEDI-MP and its application to a hypothetical accident of a nuclear submarine in the Japan Sea

    International Nuclear Information System (INIS)

    Kobayashi, Takuya; Nagai, Haruyasu; Chino, Masamichi; Togawa, Orihiko

    2004-01-01

    A software system SPEEDI-MP is being developed to resolve the environmental problems by simulating the behavior of pollutants in the atmospheric, oceanic and terrestrial environment. Verification of oceanic dispersion prediction codes on the system was carried out to assess the migration behavior of the released 241 Am from a hypothetically sunken nuclear submarine in the Japan Sea. (author)

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

    Science.gov (United States)

    Manner, Hans; Wünsch-Ziegler, Laura

    2013-08-01

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

  6. Flood control construction of Shidao Bay nuclear power plant and safety analysis for hypothetical accident of HTR-PM

    International Nuclear Information System (INIS)

    Chen Yongrong; Zhang Keke; Zhu Li

    2014-01-01

    A series of events triggered by tsunami eventually led to the Fukushima nuclear accident. For drawing lessons from the nuclear accident and applying to Shidao Bay nuclear power plant flood control construction, we compare with the state laws and regulations, and prove the design of Shidao Bay nuclear power plant flood construction. Through introducing the history of domestic tsunamis and the national researches before and after the Fukushima nuclear accident, we expound the tsunami hazards of Shidao Bay nuclear power plant. In addition, in order to verify the safety of HTR-PM, we anticipate the contingent accidents after ''superposition event of earthquake and extreme flood'', and analyse the abilities and measures of HTR-PM to deal with these beyond design basis accidents (BDBA). (author)

  7. 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)

  8. 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)

  9. 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

  10. Guidelines for calculating radiation doses to the public from a release of airborne radioactive material under hypothetical accident conditions in nuclear reactors

    International Nuclear Information System (INIS)

    1991-04-01

    This standard provides guidelines and a methodology for calculating effective doses and thyroid doses to people (either individually or collectively) in the path of airborne radioactive material released from a nuclear facility following a hypothetical accident. The radionuclides considered are those associated with substances having the greatest potential for becoming airborne in reactor accidents: tritium (HTO), noble gases and their daughters, radioiodines, and certain radioactive particulates (Cs, Ru, Sr, Te). The standard focuses on the calculation of radiation doses for external exposures from radioactive material in the cloud; internal exposures for inhalation of radioactive material in the cloud and skin penetration of tritium; and external exposures from radionuclides deposited on the ground. It uses as modified Gaussian plume model to evaluate the time-integrated concentration downwind. (52 refs., 12 tabs., 21 figs.)

  11. Comparative analysis of a hypothetical coolant loss accident in an LMFB reactor with the use of various calculation models for a common reference problem

    International Nuclear Information System (INIS)

    Royl, P.

    1979-01-01

    The results of a comparative analysis of the initial and dismantling stages of a hypothetical loss of flow accident in an LMFB reactor are presented. The analyses were made for a common reference problem with four different calculation models (CARMEN/KADIS, SURDYN, CAPRI/KADIS and FRAX). The reference core is described specifically, as are the differences in its geometrical disposition in the models, the static and transient conditions before and after the start of boiling and during dismantling. The differences in the models used for simulating the boiling and the dismantling are compared. The structure of the core, as well as the calculation conditions and hypotheses, were intentionally designed so that the accident would culminate, in all cases, in an energetic hydrodynamic dismantling stage

  12. 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

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

    International Nuclear Information System (INIS)

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

    2008-01-01

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

  14. Design features of ACR in severe accident mitigation

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Impacts on the marine environment in the case of a hypothetical accident involving the recovery of the dumped Russian submarine K-27, based on dispersion of 137Cs.

    Science.gov (United States)

    Hosseini, A; Amundsen, I; Brown, J; Dowdall, M; Karcher, M; Kauker, F; Schnur, R

    2017-02-01

    There is increasing concern regarding the issue of dumped nuclear waste in the Arctic Seas and in particular dumped objects with Spent Nuclear Fuel (SNF). Amongst dumped objects in the Arctic, the dumped Russian submarine K-27 has received great attention as it contains two reactors with highly enriched fuel and lies at a depth of about 30 m under water. To address these concerns a health and environmental impact assessment has been undertaken. Marine dispersion of potentially released radionuclides as a consequence of different hypothetical accident scenarios was modelled using the model NAOSIM. The outputs from the dispersion modelling have been used as inputs to food-chain transfer and environmental dosimetry models. The annual effective doses for subsistence fishing communities of the Barents-Kara seas region do not exceed 0.6 mSv for hypothetical accidents located at Stepovogo fjord or the Barents Sea. For high rate consumers of fish in Norway, following a potential accident at the Gremikha Bay, annual effects doses would be at around 0.15 mSv. Accumulated doses (over 90 days) for various organisms and for all release scenarios considered were never in excess of 150 μGy. The levels of 137 Cs derived for marine organism in areas close to Norway were not values that would likely cause concern from a regulatory perspective although for subsistence fishing communities close to the considered accident locations, it is not inconceivable that some restrictions on fishing etc. would need to be introduced. Copyright © 2016 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

    Llory, M.

    1988-01-01

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

  17. Development of the simulation system {open_quotes}IMPACT{close_quotes} for analysis of nuclear power plant severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi [Nuclear Power Corp. (Japan)] [and others

    1997-07-01

    The Nuclear Power Engineering Corporation (NUPEC) has initiated a long-term program to develop the simulation system {open_quotes}IMPACT{close_quotes} for analysis of hypothetical severe accidents in nuclear power plants. IMPACT employs advanced methods of physical modeling and numerical computation, and can simulate a wide spectrum of senarios ranging from normal operation to hypothetical, beyond-design-basis-accident events. Designed as a large-scale system of interconnected, hierarchical modules, IMPACT`s distinguishing features include mechanistic models based on first principles and high speed simulation on parallel processing computers. The present plan is a ten-year program starting from 1993, consisting of the initial one-year of preparatory work followed by three technical phases: Phase-1 for development of a prototype system; Phase-2 for completion of the simulation system, incorporating new achievements from basic studies; and Phase-3 for refinement through extensive verification and validation against test results and available real plant data.

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

    International Nuclear Information System (INIS)

    Shaw, Roger

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

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

  20. A framework for assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

    1991-01-01

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  2. Design consideration on severe accident for future LWR

    International Nuclear Information System (INIS)

    Omoto, A.

    1998-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-12-15

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

  4. Plant specific severe accident management - the implementation phase

    International Nuclear Information System (INIS)

    Prior, R.

    1999-01-01

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

  5. 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

  6. 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

  7. 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

  8. Formulating the Canadian regulatory position on severe accidents

    International Nuclear Information System (INIS)

    Viktorov, Alex

    2006-01-01

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

  9. 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.

  10. 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

  11. 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.

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

    International Nuclear Information System (INIS)

    Colin, M.

    1997-01-01

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

  13. Contribution of the Exposure Pathways After a Severe Accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

    Vidard, Michel

    1997-01-01

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

  15. Swedish approach to information needs in severe accident situations

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-07-01

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

  16. Swedish approach to information needs in severe accident situations

    International Nuclear Information System (INIS)

    Soederman, E.; Karnik, P.

    1992-01-01

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

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

    International Nuclear Information System (INIS)

    Gonzalez Gonzalez, F.; Jimenez Fernandez, A.

    1995-01-01

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

  18. Accident resistant transport container

    Science.gov (United States)

    Anderson, J.A.; Cole, K.K.

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident.

  19. Accident resistant transport container

    International Nuclear Information System (INIS)

    Andersen, J.A.; Cole, J.K.

    1980-01-01

    The invention relates to a container for the safe air transport of plutonium having several intermediate wood layers and a load spreader intermediate an inner container and an outer shell for mitigation of shock during a hypothetical accident

  20. 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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-10-15

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

  2. 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)

  3. 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

  4. 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

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

    Science.gov (United States)

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

    2006-01-01

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

  6. 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.

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

    International Nuclear Information System (INIS)

    Nitheanandan, Thambiayah

    2012-01-01

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

  8. 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

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

    International Nuclear Information System (INIS)

    Kovacs, Zoltan; Rydzi, Stanislav

    2015-01-01

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

  10. 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.

  11. Mitigation of severe accidents in Swedish nuclear power plants

    International Nuclear Information System (INIS)

    Soederman, E.

    1987-01-01

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

  12. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

    Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

    2001-01-01

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

  13. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  15. 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.

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

    International Nuclear Information System (INIS)

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

    1990-12-01

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US report in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Grand Gulf Nuclear Station, Unit 1. This power plant, located in Port Gibson, Mississippi, is operated by the System Energy Resources, Inc. (SERI). The emphasis in this risk analysis was not on determining a ''so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events internal to the power plant was assessed. This document provides Appendices A through E for this report. Topics included are, respectively: supporting information for the accident progression analysis; supporting information for the source term analysis; supporting information for the consequence analysis; risk results; and sampling information

  17. ANS severe accident program overview ampersand planning document

    International Nuclear Information System (INIS)

    Taleyarkhan, R.P.

    1995-09-01

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

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

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

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

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

    International Nuclear Information System (INIS)

    Majumdar, S.

    1996-09-01

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

  20. 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

  1. 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)

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

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  3. MELCOR Severe Accident Analysis on the SMART Reactor

    International Nuclear Information System (INIS)

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

    2001-01-01

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

  5. Dose to man from a hypothetical loss-of-coolant accident at the Rancho Seco Nuclear Power Plant

    International Nuclear Information System (INIS)

    Peterson, K.R.; Greenly, G.D.

    1981-02-01

    At the request of the Sacramento Municipal Utilities District, we used our computer codes, MATHEW and ADPIC, to assess the environmental impact of a loss-of-coolant accident at the Rancho Seco Nuclear Power Plant, about 40 kilometres southeast of Sacramento, California. Meteorological input was selected so that the effluent released by the accident would be transported over the Sacramento metropolitan area. With the release rates provided by the Sacramento Municipal Utilities District, we calculated the largest total dose for a 24-hour release as 70 rem about one kilometre northwest of the reactor. The largest total dose in the Sacramento metropolitan area is 780 millirem. Both doses are from iodine-131, via the forage-cow-milk pathway to an infant's thyroid. The largest dose near the nuclear plant can be minimized by replacing contaminated milk and by giving the cows dry feed. To our knowledge, there are no milk cows within the Sacramento metropolitan area

  6. Hypothetical accident scenario analyses for a 250-MW(t) modular high temperature gas-cooled reactor

    International Nuclear Information System (INIS)

    Harrington, R.M.; Ball, S.J.; Cleveland, J.C.

    1985-11-01

    This paper describes calculations performed to characterize the inherent safety of a 250-MW(t), 100-MW(e), pebble bed modular high temperature gas-cooled reactor (HTGR) design with vertical in-line arrangement (i.e., upflow core with steam generators directly above the core). A variety of postulated accident sequences involving combinations of loss of forced primary coolant (helium) circulation, loss of primary coolant pressurization, and loss of heat sink were studied and were discussed

  7. 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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1995-07-01

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

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

    International Nuclear Information System (INIS)

    1995-01-01

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

  10. 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

  11. Pending issues for severe accident management in Wolsong plants

    International Nuclear Information System (INIS)

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

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

    1987-02-01

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

  15. 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)

  16. 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

  17. 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

  18. 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

  19. Determination of doses to different organs and prediction of health detriment, after hypothetical accident in mtr reactor core

    Energy Technology Data Exchange (ETDEWEB)

    Amin, E A; Abd El-Ghani, A H [National Center of Nuclear Safety and Radiation Control Atomic Energy Authority, Cairo (Egypt)

    1997-12-31

    As a result of pypothetical accidents with release of high amount of fission products, the doses to different organs consequent upon inhalation of radioactive fission products are calculated. The processes are modeled using the ORIGIN and TIRION-4 codes: source term, containment and activity enclosure, time dependent activity behaviour in the building, and radiation exposure in the reactor building. Prediction of health detriments were calculated using ICRP-60 nominal probability coefficients and organ doses determined for bone, lung, and thyroid gland, after whole body exposure from internal inhalation and external emmersion. 11 tabs.

  20. 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)

  1. 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.

  2. Iodine chemistry effect on source term assessments. A MELCOR 186 YT study of a PWR severe accident sequence

    International Nuclear Information System (INIS)

    Herranz, Luis E.; Garcia, Monica; Otero, Bernadette

    2009-01-01

    Level-2 Probabilistic Safety Analysis has demonstrated to be a powerful tool to give insights into multiple aspects concerning severe accidents: phenomena with the greatest potential to lead to containment failure, safety systems performance and, even, to identify any additional accident management that could mitigate the consequences of such an even, etc. A major result of level-2 PSA is iodine content in Source Term since it is the main responsible for the radiological impact during the first few days after a hypothetical severe accident. Iodine chemistry is known to considerably affect iodine behavior and although understanding has improved substantially since the early 90's, a thorough understanding is still missing and most PSA studies do not address it when assessing severe accident scenarios. This paper emphasizes the quantitative and qualitative significance of considering iodine chemistry in level-2 PSA estimates. To do so a cold leg break, low pressure severe accident sequence of an actual pressurized water reactor has been analyzed with the MELCOR 1.8.6 YT code. Two sets of calculations, with and without chemistry, have been carried out and compared. The study shows that iodine chemistry could result in an iodine release to environment about twice higher, most of which would consist of around 60% of iodine in gaseous form. From these results it is concluded that exploratory studies on the potential effect of iodine chemistry on source term estimates should be carried out. (author)

  3. ESTER: a new approach in modelling severe accidents

    International Nuclear Information System (INIS)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

    Figueras, J. M.

    2000-01-01

    The paper describes those most relevant aspects of research programmes and projects, on the behavior of vessel during severe accidents with partial or total reactor core fusion, performed during the last twenty years or still on-going projects, by countries or international organizations in the nuclear community, presenting the most important technical aspects, in particular the results achieved, as well as the financial and organisational aspects. The paper concludes that, throughout a joint effort of the international nuclear community, in which Spain has been present via private and public organizations, actually exist a reasonable technical and experimental knowledge of the vessel in case of severe accidents, but still there are aspects not fully solved which are the basis for continuing some programmes and for proposal of new ones. (Author)

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

    International Nuclear Information System (INIS)

    Castro Lopez, Fernando; Orden Martinez, Alfredo

    1998-01-01

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

  6. 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)

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

    International Nuclear Information System (INIS)

    Sinka, D.; Mikulicic, V.

    2000-01-01

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

  8. MAAP - modular program for analyses of severe accidents

    International Nuclear Information System (INIS)

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

    1990-01-01

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

  9. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  10. 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

  11. 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

  12. Analyses of hypothetical FCI's in a fast reactor

    International Nuclear Information System (INIS)

    Padilla, A. Jr.; Martin, F.J.; Niccoli, L.G.

    1981-01-01

    Parametric analyses using the SIMMER code were performed to evaluate the potential for a severe recriticality from a pressure-driven recompaction caused by an energetic FCI during the transition phase of a hypothetical accident in a fast reactor. For realistic and reasonable estimates for the assumed accident conditions, a severe recriticality was not predicted. The conditions under which a severe recriticality would be obtained or averted were identified. 10 figures, 2 tables

  13. 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)

  14. 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

  15. 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.

  16. Influence of radiation heat transfer during a severe accident

    International Nuclear Information System (INIS)

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

    2016-09-01

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

  17. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

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

    1990-08-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  19. 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)

  20. Health effects of technologies for power generation: Contributions from normal operation, severe accidents and terrorist threat

    International Nuclear Information System (INIS)

    Hirschberg, Stefan; Bauer, Christian; Burgherr, Peter; Cazzoli, Eric; Heck, Thomas; Spada, Matteo; Treyer, Karin

    2016-01-01

    As a part of comprehensive analysis of current and future energy systems we carried out numerous analyses of health effects of a wide spectrum of electricity supply technologies including advanced ones, operating in various countries under different conditions. The scope of the analysis covers full energy chains, i.e. fossil, nuclear and renewable power plants and the various stages of fuel cycles. State-of-the-art methods are used for the estimation of health effects. This paper addresses health effects in terms of reduced life expectancy in the context of normal operation as well as fatalities resulting from severe accidents and potential terrorist attacks. Based on the numerical results and identified patterns a comparative perspective on health effects associated with various electricity generation technologies and fuel cycles is provided. In particular the estimates of health risks from normal operation can be compared with those resulting from severe accidents and hypothetical terrorist attacks. A novel approach to the analysis of terrorist threat against energy infrastructure was developed, implemented and applied to selected energy facilities in various locations. Finally, major limitations of the current approach are identified and recommendations for further work are given. - Highlights: • We provide state-of-the-art comparative assessment of energy health risks. • The scope of the analysis should to the extent possible cover full energy chains. • Health impacts from normal operation dominate the risks. • We present novel approach to analysis of terrorist threat. • Limitations include technology choices, geographical coverage and terrorist issues.

  1. 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.

  2. 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)

  3. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-15

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

  4. Severe accident development modeling and evaluation for CANDU

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  5. 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

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

    Directory of Open Access Journals (Sweden)

    Andgren Karin

    2017-01-01

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

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

    Science.gov (United States)

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

    2014-05-01

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

  8. SARNET integrated European Severe Accident Research-Conclusions in the source term area

    Energy Technology Data Exchange (ETDEWEB)

    Haste, T., E-mail: tim.haste@irsn.f [Paul Scherrer Institute, CH-5232 Villigen PSI (Switzerland); Giordano, P. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Herranz, L. [Centro de Investigaciones Energeticas Medio Ambientales y Tecnologica, CIEMAT, Avda. Complutense 22, E-28040 Madrid (Spain); Girault, N.; Dubourg, R. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Sabroux, J.-C. [Institut de Radioprotection et de Surete Nucleaire, IRSN, Saclay Research Centre, BP 68, F-91192 Gif-sur-Yvette Cedex (France); Cantrel, L. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France); Bottomley, D. [European Commission Joint Research Centre, Transuranium Institute, P.O. Box 2340, D-76125 Karlsruhe (Germany); Parozzi, F. [ENEA - Ricerca sul Sistema Elettrico (ERSE) SpA., Via Rubattino 54, I-20134 Milano (Italy); Auvinen, A. [VTT Technical Research Centre of Finland, P.O. Box 1000, FI-02044 VTT Espoo (Finland); Dickinson, S. [National Nuclear Laboratory, Harwell Business Centre, Didcot, OX11 0QJ (United Kingdom); Lamy, J.-C. [Electricite de France, 12-14 avenue Dutrievoz, F-69100 Villeurbanne (France); Weber, G. [Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbH, Forschungsgelaende, D-85748 Garching (Germany); Albiol, T. [Institut de Radioprotection et de Surete Nucleaire, IRSN, BP 3, F-13115 St Paul lez Durance Cedex (France)

    2009-12-15

    The overall aim of the SARNET (Severe Accident Research NETwork), in the EU 6th Framework programme was to integrate in a sustainable manner the research capabilities of fifty-one European organisations from eighteen member states of the European Union (EU) plus the Joint Research Centres, with one Canadian company, to resolve important remaining uncertainties and safety issues concerning existing and future nuclear plant, especially water-cooled reactors, under hypothetical severe accident conditions. It emphasised integrating activities, spreading of excellence (including knowledge transfer) and jointly executed research, with the knowledge gained being encapsulated in the European severe accident modelling code ASTEC. This paper summarises the achievements over the whole project in the Source Term Topic, which dealt with potential radioactive release to the environment, covering release of fission products and structural materials from the core, their transport in the primary circuit, and their behaviour in the containment. The main technical areas covered, as emphasised by the earlier EURSAFE project, were the effect of oxidative conditions on fission product release and transport (especially the behaviour of the highly radiotoxic ruthenium under air ingress conditions), iodine volatility in the primary circuit, control rod aerosol release (Ag-In-Cd) that affects iodine transport, containment by-pass in the case of steam generator tube rupture, aerosol retention in containment cracks, aerosol remobilisation in the circuit, and iodine/ruthenium behaviour in the containment especially concerning the volatile fraction in the atmosphere. The studies also covered performance of new experiments, analysis of existing data, and formulation and improvement of theoretical models. Significant progress was made in each area. Looking to the future, the 7th Framework successor project SARNET2 covers the remaining issues concerning iodine and ruthenium, including practical

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  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. Millstone Unit 1 plant vulnerabilities during postulated severe nuclear accidents

    International Nuclear Information System (INIS)

    Khalil, Y.F.

    1993-01-01

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

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

    Directory of Open Access Journals (Sweden)

    Gilberto Espinosa-Paredes

    2012-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-15

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy.

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

    International Nuclear Information System (INIS)

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

    1991-12-01

    The objectives of this study is to develop a system of computer codes for postulated severe accident analyses in Nuclear Power Plants. This system of codes is necessary to conduct individual plant examination for domestic nuclear power plants. As a result of this study, one can conduct severe accident assessments more easily, and can extract the plant-specific vulnerabilities for severe accidents and at the same time the ideas for enhancing overall accident resistance. The scope and contents of this study are as follows : development of a system of computer codes for severe accident analyses, development of severe accident management strategy

  15. 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

  16. Modeling of severe accident sequences with the new modules CESAR and DIVA of ASTEC system code

    International Nuclear Information System (INIS)

    Pignet, Sophie; Guillard, Gaetan; Barre, Francois; Repetto, Georges

    2003-01-01

    Systems of computer codes, so-called 'integral' codes, are being developed to simulate the scenario of a hypothetical severe accident in a light water reactor, from the initial event until the possible radiological release of fission products out of the containment. They couple the predominant physical phenomena that occur in the different reactor zones and simulate the actuation of safety systems by procedures and by operators. 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 should take less than one day of real time to simulate on a PC computer. This search of compromise is a real challenge for such integral codes. The development of the ASTEC integral code was initiated jointly by IRSN and GRS as an international reference code. The latest version 1.0 of ASTEC, including the new modules CESAR and DIVA which model the behaviour of the reactor cooling system and the core degradation, is presented here. Validation of the modules and one plant application are described

  17. Status of the French R/D program on the severe accident issue to develop Gen IV SFRs - 15373

    International Nuclear Information System (INIS)

    Serre, F.; Bertrand, F.; Journeau, C.; Suteau, C.; Verwaede, D.; Schmitt, D.; Farges, B.

    2015-01-01

    The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are to prevent core melting, in particular by the development of an innovative core with complementary safety prevention devices, and to enhance the reactor resistance to severe accident by design. To mitigate the consequences of hypothetical core melting situations, specific dispositions or mitigation devices will be added to the core and to the reactor. It is also required to provide a robust safety demonstration (with high level of confidence). Therefore a new approach for severe accident issue has been defined: to the well-known 'lines of defense' method, a 'lines of mitigation' method is added. To meet these ASTRID, or future SFR, requirements, a large R/D program was launched in the Severe Accident domain, with a large number of partners. This paper will present the status of the CEA R/D related to the SFR Severe Accident issue, the collaboration framework (with industrial partners and R/D foreign organizations), and the future R/D plans to support the ASTRID project and possible developments for future Gen IV commercial SFR. (authors)

  18. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

    Sehga, B.R.

    1998-01-01

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

  19. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

    Nichols, A.L.

    1990-09-01

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

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

    International Nuclear Information System (INIS)

    Fontana, M.H.

    1980-10-01

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

  1. Centrifugal Filtration System for Severe Accident Source Term Treatment

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  2. 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

  3. 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)

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

    International Nuclear Information System (INIS)

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

    1991-03-01

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

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

    International Nuclear Information System (INIS)

    Holtzclaw, K.W.

    1983-01-01

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

  6. 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

  7. Numerical Study of Severe Accidents on Containment Venting Conditions

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  8. 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

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

    International Nuclear Information System (INIS)

    Jeong, Kwang Il; Lee, Joon Ku

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  11. 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)

  12. Most likely failure location during severe accident conditions

    International Nuclear Information System (INIS)

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

    1991-01-01

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

  13. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  15. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-01

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

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

    International Nuclear Information System (INIS)

    Hill, Axel; Stiepani, Cristoph; Drechsler, Michael

    2015-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP either due to containment leakages or due to intentional filtered containment venting. In the latter case aerosols and iodine are retained, however noble gases are not retainable by the FCVS or by conventional air filtration systems like HEPA filters and iodine absorbers. Radioactive noble gases nevertheless dominate the activity release depending on the venting procedure and the weather conditions. To prevent unacceptable contamination of the control room atmosphere by noble gases, AREVA GmbH has developed a noble gas control room accident filtration system (CRAFT) which can supply purified fresh air to the control room without time limitation. The retention process is based on dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. CRAFT allows minimization of the dose rate inside the control room and ensures low radiation exposure to the staff by maintaining the control room environment suitable for prolonged occupancy throughout the duration of the accident. CRAFT consists of a proven modular design either transportable or permanently installed. (author)

  17. 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

  18. Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

    International Nuclear Information System (INIS)

    Matsuki, Yoshio; Muramatsu, Ken

    2008-05-01

    The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

  19. 77 FR 66649 - Proposed Revision to Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-11-06

    ... and Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION... Assessment and Severe Accident Evaluation for New Reactors.'' The NRC is extending the public comment period... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  20. 77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

    Science.gov (United States)

    2012-10-09

    ... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

  1. Noble gas control room accident filtration system for severe accident conditions N-CRAFT. System design

    International Nuclear Information System (INIS)

    Hill, Axel

    2014-01-01

    Severe accidents might cause the release of airborne radioactive substances to the environment of the NPP. This can either be due to leakages of the containment or due to a filtered containment venting in order to ensure the overall integrity of the containment. During the containment venting process aerosols and iodine can be retained by the FCVS which prevents long term ground contamination. Noble gases are not retainable by the FCVS. From this it follows that a large amount of radioactive noble gases (e.g. xenon, krypton) might be present in the nearby environment of the plant dominating the activity release, depending on the venting procedure and the weather conditions. Accident management measures are necessary in case of severe accidents and the prolonged stay of staff inside the main control room (MCR) or emergency response center (ERC) is essential. Therefore, the in leakage and contamination of the MRC and ERC with airborne activity has to be prevented. The radiation exposure of the crises team needs to be minimized. The entrance of noble gases cannot be sufficiently prevented by the conventional air filtration systems such as HEPA filters and iodine absorbers. With the objective to prevent an unacceptable contamination of the MCR/ERC atmosphere by noble gases AREVA GmbH has developed a noble gas retention system. The noble gas control room accident filtration system CRAFT is designed for this case and provides supply of fresh air to the MCR/ERC without time limitation. The retention process of the system is based on the dynamic adsorption of noble gases on activated carbon. The system consists of delay lines (carbon columns) which are operated by a continuous and simultaneous adsorption and desorption process. These cycles ensure a periodic load and flushing of the delay lines retaining the noble gases from entering the MCR. CRAFT allows a minimization of the dose rate inside MCR/ERC and ensures a low radiation exposure to the staff on shift maintaining

  2. Theoretical and experimental investigations on the behaviour of iodine during severe accidents: volatile iodine. Final report

    International Nuclear Information System (INIS)

    Funke, F.; Zeh, P.; Greger, G.U.; Hellmann, S.

    1999-01-01

    Analysis of the consequences of severe accidents in nuclear power plants requires knowledge of the behaviour of radionuclides relevant from the radiological viewpoint, especially the iodine. The current modelling of iodine behaviour is not conclusive, owing to insufficiently known data. This project is intended to eliminate some of these data gaps in critical areas. 350 tests on the radiation-induced oxidation of elemental iodine (I 2 ) in the containment atmosphere were performed yielding an extended database. Moreover, irradiation tests were performed on the formation and decomposition of ozone which is a reaction partner for I 2 . The reaction with ozone converts volatile I 2 into non-volatile iodine oxides or iodate. An improved kinetic modelling was developed for the iodine accident code IMPAIR. Now the model is valid also for steam-containing atmospheres and, additionally, considers dose rate and thus the actual ozone concentration. An assessment of the literature concludes that β and γ radiation have no different impact on iodine chemistry and thus do not need to be modelled separately in iodine accident codes. An assessment of the literature shows a partly significant chemical interaction of volatile iodine with aerosols. Since such reactions lead to a faster decrease of volatile iodine at least at high aerosol concentrations, a modelling should be foreseen in the future. In the frame of the international ISP-41 project, calculations to an integral test in the Canadian Radioiodine Test Facility (RTF) were performed with IMPAIR. The existing model of the radiation-induced I 2 formation in the sump in IMPAIR is identified as a weakness requiring future improvement. A theoretical assessment on the iodine chemistry in the droplets of a spray system concludes that a modelling is necessary in case of spraying with fresh water, and that this is already contained in available spray models. During recirculation spraying in an examplary, hypothetical EPR case, no

  3. Analysis of hypothetical incidents in nuclear power plants with PWR and HTR

    International Nuclear Information System (INIS)

    Geiser, H.

    1977-01-01

    Several accident analyses are reviewed with a view to fission product release, and the findings are transferred to German reactor plants with LWR and HTR and compared. First of all, hypothetical accidents are compared for both of these lines; after this, the history of accidents is briefly described, and the fission product release during these accidents is investigated. For both reactor lines, there is a different but sufficiently high potential for safety improvements. (orig.) [de

  4. 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

  5. Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident

    International Nuclear Information System (INIS)

    Journeau, Ch.

    2008-01-01

    The PLINIUS experimental platform at CEA Cadarache is dedicated to the experimental study of nuclear reactor severe accidents thanks to experiments between 2000 and 3500 K with prototypic corium. Corium is the mixture that would be formed by an hypothetical core melting and its mixing with structural materials. Prototypical corium has the same chemical composition as the corium corresponding to a given accident scenario but has a different isotopic composition (use of depleted uranium,...). Research programs and test series have been performed to study corium thermophysical properties, fission product behaviour, corium spreading, solidification and interaction with concrete as well as its coolability. It was the frame of research training of many students and was realized within national, European and international collaborations. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

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

    International Nuclear Information System (INIS)

    Jang, H. S.; Jeon, M. H.; Cho, N. J. and others

    1992-01-01

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

  8. Application of the severe accident code ATHLET-CD. Modelling and evaluation of accident management measures (Project WASA-BOSS)

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

    The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. Numerical analyses are used to investigate the accident progression and the complex physical phenomena during the core degradation phase, as well as to evaluate the effectiveness of possible countermeasures in the preventive and mitigative domain [1, 2]. The presented analyses have been performed with the computer code ATHLET-CD developed by GRS [3, 4].

  9. Fukushima-Daiichi after the severe accident (estimation)

    International Nuclear Information System (INIS)

    Otcenasek, Petr

    2011-01-01

    All facts about the Fukushima-Daiichi NPP and about the accident known at the time of publication are summarized and expected remedial actions and consequences of the accidents are deduced. The paper is structured as follows: (1) Accident initiation is known; (2) Logically inferred results; (3) Framework identification; (4) Survey; and (5) Economic and strategic impacts of the accident. Worldwide solidarity is mentioned in conclusion. (P.A.)

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

    International Nuclear Information System (INIS)

    Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

    1992-03-01

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

  11. 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.

  12. Integrated severe accident containment analysis with the CONTAIN computer code

    International Nuclear Information System (INIS)

    Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

    1985-12-01

    Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

  13. 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

  14. Development of severe accident management guidance for Younggwang units 5 and 6

    International Nuclear Information System (INIS)

    Lee, K. W.; Beon, C. S.; Kim, M. K.; Hong, S. Y.; Park, K. S.

    2001-01-01

    Severe Accident Management Guidance (SAMG) has been developed for Younggwang Units 5 and 6. It is consisted of Severe Accident Control Room Guideline, Diagnostic Flow Chart, Severe Accident Guideline, Severe Challenge Guideline, TSC Long Term Monitoring, SAMG Termination. Severe Accident Control Room Guideline, which deals with severe accident after finishing Emergency Operation Procedure, consists of acitions before and after TSC actuation. Seven servere accident management strategies are developed. Diagnostic Flow Chart, Severe Accident Guideline, and Severe Challenge Guideline are developed for each strategy, which enables the users to the implementation of strategy easily and systematically. TSC Long Term Monitoring is also developed to monitor long term activities after a particular strategy. Total of 45 set points are developed for decision making during the implementation of the SAMG

  15. A study on the hydrogen behavior and its mitigation in the APR1400 containment during a severe accident

    International Nuclear Information System (INIS)

    Kim, Jong Tae; Hong, Seong Wan; Park, Rae Joon; Kim, Sang Baik

    2005-02-01

    During a hypothetical severe accident in a nuclear power plant (NPP), hydrogen is generated by the active reaction of fuel-cladding and steam in the reactor pressure vessel and released with steam into the containment. In this study, the analysis of the hydrogen and steam behavior during selected severe accidents in the APR1400 containment has been conducted by using the GASFLOW code. For the SBLOCA, hydrogen was accumulated in the containment dome region quickly when only PARSs were used. When the igniters were turned on, a standing flame was formed around a coolant pump and burnt most of the hydrogen blown-out. For the TLOFW accident, the flap-type pressure damper installed at the IRWST vents strongly affected the flow structure of the hydrogen. And by the steam-rich and oxygen starvation conditions in the IRWST, DDT is not likely to occur. For the SBO accident, dry hydrogen was release in the IRWST by the assumption of full condensation of the released steam in the IRWST water. In this case, the possibility of flame acceleration is high in the IRWST and annular compartment. In this study two design modifications were proposed in view of the hydrogen mitigation strategy and their effectiveness was evaluated by the GASFLOW analysis

  16. 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