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

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.

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

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

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

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

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

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

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)

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

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

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

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

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

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)

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

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

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

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

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)

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

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)

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.)

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

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.

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

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)

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

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

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

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.

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

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

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

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)

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)

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

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

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

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

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

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)

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

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.

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

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

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

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)

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

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)

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

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

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

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.)

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

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)

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

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)

Filtra-MVSS mitigates the consequences of a severe reactor accident

International Nuclear Information System (INIS)

Lindau, L.; Gustavsson, L.; Elisson, K.

1989-01-01

The purpose of the filter system, known as Filtra-MVSS (for Multi-Venturi Scrubber System) is to provide pressure relief for the reactor containment and to limit radioactive releases in the event of a reactor accident, thereby largely preventing local radioactive fall-out. To this end the pressure is reduced in a controlled manner and the discharged steam and gases are washed in a multi-venturi scrubber before release. The system is flexible and can cope with a number of hypothetical accident scenarios

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

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.

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.

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)

Analysis of radionuclide behavior in a BWR Mark-II containment under severe accident management condition in low pressure sequence

International Nuclear Information System (INIS)

Funayama, Kyoko; Kajimoto, Mitsuhiro; Nagayoshi, Takuji; Tanaka, Nobuo

1999-01-01

In the Level 2 PSA program at INS/NUPEC, MELCOR1.8.3 is extensively applied to analyze radionuclide behavior of dominant sequences. In addition, the revised source terms provided in the NUREG-1465 report have been also discussed to examine the potential of the radionuclides release to the environment in the conventional siting criteria. In the present study, characteristics of source terms to the environment were examined comparing with results by the Hypothetical Accident (LOCA), NUREG-1465 and MELCOR1.8.3. calculation for a typical BWR with a Mark-II containment in order to assure conservatives of the Hypothetical Accident in Japan. Release fractions of iodine to the environment for the Hypothetical Accident and NUREG-1465, which used engineering models for predicting radionuclide behaviors, were about 10 -4 and 10 -6 of core inventory, respectively, while the best estimate MELCOR1.8.3 code predicted 10 -9 of iodine to the environment. The present study showed that the engineering models in the Hypothetical Accident or NUREG-1465 have large conservatives to estimate source term of iodine to the environment. (author)

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)

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

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

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

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)

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.)

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.

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.

Source terms associated with two severe accident sequences in a 900 MWe PWR

International Nuclear Information System (INIS)

Fermandjian, J.; Evrard, J.M.; Berthion, Y.; Lhiaubet, G.; Lucas, M.

1983-12-01

Hypothetical accidents taken into account in PWR risk assessment result in fission product release from the fuel, transfer through the primary circuit, transfer into the reactor containment building (RCB) and finally release to the environment. The objective of this paper is to define the characteristics of the source term (noble gases, particles and volatile iodine forms) released from the reactor containment building during two dominant core-melt accident sequences: S 2 CD and TLB according to the ''Reactor Safety Study'' terminology. The reactor chosen for this study is a French 900 MWe PWR unit. The reactor building is a prestressed concrete containment with an internal liner. The first core-melt accident sequence is a 2-break loss-of-coolant accident on the cold leg, with failure of both system and the containment spray system. The second one is a transient initiated by a loss of offsite and onsite power supply and auxiliary feedwater system. These two sequences have been chosen because they are representative of risk dominant scenarios. Source terms associated with hypothetical core-melt accidents S 2 CD and TLB in a French PWR -900 MWe- have been performed using French computer codes (in particular, JERICHO Code for containment response analysis and AEROSOLS/31 for aerosol behavior in the containment)

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)

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

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

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)

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

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

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

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.

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.

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)

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)

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

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)

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

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.)

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.

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

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.

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

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.

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

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)

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.

Filtra-MVSS mitigates the consequences of a severe reactor accident

International Nuclear Information System (INIS)

Lindau, L.; Gustavsson, L.; Elisson, K.

1989-01-01

ABB Atom and Flaekt Industri AB, both members of the ABB Group, have jointly developed an efficient filter for use in nuclear power stations. The purpose of the filter system, known as Filtra-MVSS (for Multi-Venturi Scrubber System) is to provide pressure relief for the reactor containment and to limit radioactive releases in the event of a reactor accident, thereby largely preventing local radioactive fall-out. To this end the pressure is reduced in a controlled manner and the discharged steam and gases are washed in multi-venturi scrubber before release. The system is flexible and can cope with a number of hypothetical accident scenarios. (orig.) [de

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.)

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)

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

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)

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

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

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

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).

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

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

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

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.)

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

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)

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

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.

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

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

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

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).

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).

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)

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

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

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)

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

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

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)

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)

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

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)

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)

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

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

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

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)

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

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.

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.)

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.)

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

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)

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

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.

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)

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

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

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report

International Nuclear Information System (INIS)

Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

2016-09-01

The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

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)

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

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

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

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...

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)

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)

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

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

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)

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

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

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

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.

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

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)

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)

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)

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)

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

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

1974-01-01

The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

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)

New system technologies implemented at Kozloduy 3 and 4 (WWER 440-230) for containment leakage and H2 control in severe accident situations - Design, qualification, installation, commissioning

International Nuclear Information System (INIS)

Feuerbach, R.; Eckardt, B.; Kastner, B.

2005-01-01

In order to reduce the residual risk associated with hypothetical severe nuclear accidents, systems and components for filtered containment venting and H 2 reduction were developed. During severe accident scenarios large quantities of hydrogen and radioactive material may be released into the containment atmosphere within a short period of time. In the event of internal over pressurization due to hypothetical severe accident sequences a pressure barrier system has to be created to confine the activity in the containment. Unavoidable releases of activity to the environment have to be minimized to a great extent as possible. Research into the hypothetical event of core melt accidents has continued and new accident mitigation technologies have been developed. Decisions have been taken to implement these new mitigation measures in operating nuclear power plants to mitigate severe accidents consequences. In order to prevent loss of containment integrity as a result of over pressurization, nuclear power plants in the Federal Republic of Germany as well as in most other European countries have been or will be back-fitted with systems for filtered venting of the containment atmosphere and systems for H 2 -control. Similar technologies for containment venting system and H 2 control have been now implemented in the first WWER 440-230 units of Kozloduy 3 and 4. Following OECD recommendations sever accident situations were analyzed and a design of countermeasures have been performed. Main goal of the developed countermeasures was to overcome the WWER 440-230 containment design specifics like, leakage rate behavior, limited available containment volume combined with the feature of high availability of electrical supply at multiple plant sites. Further more the design of counter measures considers the common use for Kozloduy unit 3 and 4. The analysis of postulated severs accident situation - without countermeasures - showed significant increase of H 2 /O 2 concentration in the

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)

WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject F. Contributions to code validation using BWR data and to evaluation and optimization of accident management measures. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt F. Beitraege zur Codevalidierung anhand von SWR-Daten und zur Bewertung und Optimierung von Stoerfallmassnahmen. Schlussbericht

Energy Technology Data Exchange (ETDEWEB)

Di Marcello, Valentino; Imke, Uwe; Sanchez Espinoza, Victor

2016-09-15

The exact knowledge of the transient course of events and of the dominating processes during a severe accident in a nuclear power station is a mandatory requirement to elaborate strategies and measures to minimize the radiological consequences of core melt. Two typical experiments using boiling water reactor assemblies were modelled and simulated with the severe accident simulation code ATHLET-CD. The experiments are related to the early phase of core degradation in a boiling water reactor. The results reproduce the thermal behavior and the hydrogen production due to oxidation inside the bundle until relocation of material by melting. During flooding of the overheated assembly temperatures and hydrogen oxidation are under estimated. The deviations from the experimental results can be explained by the missing model to simulate bore carbide oxidation of the control rods. On basis of a hypothetical loss of coolant accident in a typical German boiling water reactor the effectivity of flooding the partial degraded core is investigated. This measure of mitigation is efficient and prevents failure of the reactor pressure vessel if it starts before molten material is relocated into the lower plenum. Considerable amount of hydrogen is produced by oxidation of the metallic components.

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)

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...

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)

Development of a prototype graphic simulation program for severe accident training

International Nuclear Information System (INIS)

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

2000-05-01

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

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

Development of the severe accident risk information database management system SARD

International Nuclear Information System (INIS)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

Development of the severe accident risk information database management system SARD

Energy Technology Data Exchange (ETDEWEB)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

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

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.

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

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

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)

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

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

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.

Development of a prototype graphic simulation program for severe accident training

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

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.

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

A multivariate tobit analysis of highway accident-injury-severity rates.

Science.gov (United States)

Anastasopoulos, Panagiotis Ch; Shankar, Venky N; Haddock, John E; Mannering, Fred L

2012-03-01

Relatively recent research has illustrated the potential that tobit regression has in studying factors that affect vehicle accident rates (accidents per distance traveled) on specific roadway segments. Tobit regression has been used because accident rates on specific roadway segments are continuous data that are left-censored at zero (they are censored because accidents may not be observed on all roadway segments during the period over which data are collected). This censoring may arise from a number of sources, one of which being the possibility that less severe crashes may be under-reported and thus may be less likely to appear in crash databases. Traditional tobit-regression analyses have dealt with the overall accident rate (all crashes regardless of injury severity), so the issue of censoring by the severity of crashes has not been addressed. However, a tobit-regression approach that considers accident rates by injury-severity level, such as the rate of no-injury, possible injury and injury accidents per distance traveled (as opposed to all accidents regardless of injury-severity), can potentially provide new insights, and address the possibility that censoring may vary by crash-injury severity. Using five-year data from highways in Washington State, this paper estimates a multivariate tobit model of accident-injury-severity rates that addresses the possibility of differential censoring across injury-severity levels, while also accounting for the possible contemporaneous error correlation resulting from commonly shared unobserved characteristics across roadway segments. The empirical results show that the multivariate tobit model outperforms its univariate counterpart, is practically equivalent to the multivariate negative binomial model, and has the potential to provide a fuller understanding of the factors determining accident-injury-severity rates on specific roadway segments. Published by Elsevier Ltd.

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

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)

Melt Fragmentation Characteristics of Metal Fuel with Melt Injection Mass during Initiating Phase of SFR Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Heo, Hyo; Lee, Min Ho; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of); Jerng, Dong Wook [Chung-Ang Univ., Seoul (Korea, Republic of)

2016-05-15

The PGSFR has adopted the metal fuel for its inherent safety under severe accident conditions. However, this fuel type is not demonstrated clearly yet under the such severe accident conditions. Additional experiments for examining these issues should be performed to support its licensing activities. Under initiating phase of hypothetic core disruptive accident (HCDA) conditions, the molten metal could be better dispersed and fragmented into the coolant channel than in the case of using oxide fuel. This safety strategy provides negative reactivity driven by a good dispersion of melt. If the coolant channel does not sufficient coolability, the severe recriticality would occur within the core region. Thus, it is important to examine the extent of melt fragmentation. The fragmentation behaviors of melt are closely related to a formation of debris shape. Once the debris shape is formed through the fragmentation process, its coolability is determined by the porosity or thermal conductivity of the melt. There were very limited studies for transient irradiation experiments of the metal fuel. These studies were performed by Transient Reactor Test Facility (TREAT) M series tests in U.S. The TREAT M series tests provided basic information of metal fuel performance under transient conditions. The effect of melt injection mass was evaluated in terms of the fragmentation behaviors of melt. These behaviors seemed to be similar between single-pin and multi-pins failure condition. However, the more melt was agglomerated in case of multi-pins failure.

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...

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

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)

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

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.

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

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)

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)

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)

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)

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

Factors associated with urban non-fatal road-accident severity.

Science.gov (United States)

Potoglou, Dimitris; Carlucci, Fabio; Cirà, Andrea; Restaino, Marialuisa

2018-02-05

This paper reports on the factors associated with non-fatal urban-road accident severity. Data on accidents were gathered from the local traffic police in the City of Palermo, one of the six most populated cities in Italy. Findings from a mixed-effects logistic-regression model suggest that accident severity increases when two young drivers are involved, road traffic conditions are light/normal and when vehicles crash on a two-way road or carriageway. Speeding is more likely to cause slight or serious injury even when compared to a vehicle moving towards the opposite direction of traffic. An accident during the summer is more likely to result in a slight or serious injury than an accident during the winter, which is in line with evidence from Southern Europe and the Middle East. Finally, the severity of non-fatal accident injuries in an urban area of Southern Europe was significantly associated with speeding, the age of the driver and seasonality.

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)

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)

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)

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

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

Knowledge data base for severe accident management of nuclear power plants

International Nuclear Information System (INIS)

Ogino, Masao; Kawabe, Ryuhei; Nagasaka, Hideo; Sumida, Susumu; Fukasawa, Masanori; Muta, Hitoshi

2011-01-01

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

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.

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

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)

Severe accident simulation and analysis for a CAREM-like integral nuclear reactor: ex-vessel phase

International Nuclear Information System (INIS)

Caputo, M.; García, J.M.; Giménez, M.; Sánchez, S.

2013-01-01

The main phenomena and processes involved in the progression of a hypothetical nuclear severe accident in an integral type reactor like CAREM are studied, quantifying the most relevant parameters, in order to contribute to the plant design and the development of an appropriate severe accident management program. A computational plant model was developed using Melcor code, including the reactor pressure vessel and the containment. A loss of coolant accident caused by a double guillotine pipe break in the steam dome zone of the pressure vessel (1.5 inches diameter) was simulated. Along this work the analysis were focused in the containment dynamics. As a consequence of the postulated loss of coolant accident the water inventory boils off leading to the core uncovery and fuel heat-up. At high temperatures the zircaloy steam oxidation becomes relevant, with hydrogen generation as one of the reaction products. The hydrogen produced is release into the containment and the possibility of hydrogen combustion in presence of enough oxygen makes relevant the analysis of containment hydrogen distribution. It is assumed that there is not any hydrogen control system. Due to the postulated loss of coolant a big amount of steam and energy is released into the containment, with a consequent fast pressurization of the dry well which makes possible air and steam discharging into the wet well (suppression pool). At the beginning the flow discharged into the pool is mainly composed of air, a non-condensable gas that pressurizes the wet well. After most of the containment air is pushed into the atmosphere wet well the pressurization rate decreases because the flow discharge is mainly composed by steam, which condensates in the pool. Also some other containment pressure peaks were observed as a consequence of hydrogen deflagrations. (author)

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)

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)

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

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

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)

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)

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

Containment pressure monitoring method after severe accident in nuclear power plant

International Nuclear Information System (INIS)

Luo Chuanjie; Zhang Shishui

2011-01-01

The containment atmosphere monitoring system in nuclear power plant was designed on the basis of design accident. But containment pressure will increase greatly in a severe accident, and pressure instrument in the containment can't satisfy the monitoring requirement. A new method to monitor the pressure change in the containment after a severe accident was considered, through which accident soften methods can be adopted. Under present technical condition, adding a pressure monitoring channel out of containment for post-severe accident is a considerable method. Daya Bay Nuclear Power Plant implemented this modification, by which the containment release time can be delayed during severe accident, and nuclear safety can be increased. After analysis, this method is safe and feasible. (authors)

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

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)

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.

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

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

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

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

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

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)

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.)

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)

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.)

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

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.

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

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

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)

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)

Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

International Nuclear Information System (INIS)

1992-04-01

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

Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-04-15

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

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

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

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

CE/Bechtel design containment response to severe accident phenomenology: A comparison among several combustion engineering plants

International Nuclear Information System (INIS)

Khalil, Y.F.; Schneider, R.E.

1995-01-01

The objectives of this paper are to: (1) discuss the types of severe accident phenomena that drive containment failure modes in CE plants and (2) contribute to the current state of knowledge of CE/Bechtel-design containment response to severe accident phenomenology. The second objective is addressed by providing a comparative study of containment response to severe accidents among several CE plants including Millstone Unit 2 (MP2), Palisades (Consumers Power), Calvert Cliffs (Baltimore Gas and Electric Company), Palo Verde (Arizona Public Service), and SONGS Units 2 and 3 (Southern California Edison). The motivation for addressing the second objective is based on the current lack of comprehensive literature on CE/Bechtel design containment failure modes and mechanisms for accidents that progress beyond the design basis limits. The first part of this paper addresses severe accident phenomena-related failure mechanisms in CE/Bechtel-designed containments. The second part of this work provides a comparative study of containment response among several CE plants

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

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

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

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.

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

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

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

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

1998-12-31

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

Thermal Hydraulic design parameters study for severe accidents using neural networks

Energy Technology Data Exchange (ETDEWEB)

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

1997-12-31

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

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

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

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

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

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

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

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

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

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

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

1992-01-01

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

Long-term consequences for Northern Norway of a hypothetical release from the Kola nuclear power plant

International Nuclear Information System (INIS)

Howard, B.J.; Wright, S.M.; Salbu, B.; Skuterud, K.L.; Hove, K.; Loe, R.

2004-01-01

The spatial and temporal variation in radiocaesium and 90 Sr doses to two population groups of the two Northernmost counties of Norway, Troms and Finnmark, following a hypothetical accident at the Kola nuclear power plant (KNPP) have been estimated using a model implemented within a geographical information system. The hypothetical accident assumes a severe loss of coolant accident at the KNPP coincident with meteorological conditions causing significant radionuclide deposition in the two counties. External doses are estimated from ground deposition and the behaviour of the different population groups, and internal doses from predicted food product activity concentrations and dietary consumption data. Doses are predicted for reindeer keepers and other Norwegian inhabitants, taking account of existing 137 Cs and 90 Sr deposition but not including the remedial effect of any countermeasures that might be used. The predicted doses, arising mainly from radiocaesium, confirm the Arctic Monitoring and Assessment Programme assessment that residents of the Arctic are particularly vulnerable to radiocaesium contamination, which could persist for many years. External doses are predicted to be negligible compared to ingestion doses. Ingestion doses for reindeer keepers are predicted to exceed 1 mSv y -1 for several decades primarily due to their high consumption of reindeer meat. Other Norwegians would also be potentially exposed to doses exceeding 1 mSv y -1 for several years, especially if they consume many local products. Whilst reindeer production is the most important exposure pathway, freshwater fish, lamb meat, dairy products, mushrooms and berries are also significant contributors to predicted ingestion doses. Radionuclide fluxes, defined as the total output of radioactivity in food from an area for a unit time, are dominated by reindeer meat. The results show the need for an effective emergency response, with appropriate countermeasures, should an accident of the

Injury severity and seating position in accidents with German EMS helicopters.

Science.gov (United States)

Hinkelbein, Jochen; Spelten, Oliver; Neuhaus, Christopher; Hinkelbein, Mandy; Özgür, Enver; Wetsch, Wolfgang A

2013-10-01

Accident rates and fatality rates for Helicopter Emergency Medical Service (HEMS) missions have been investigated recently, but none of these studies considered the influence of the seating position in the helicopter. The aim of the present descriptive and observational study was to analyze injury severity depending on the seating position during HEMS accidents in Germany. Data from the German Federal Agency for Flight Accident Investigation was gathered for a period of 40 years (from 1970 to 2009). The seating position in the aircraft during the accident and the resulting injury severity (i.e., 1=no; 2=slight; 3=severe; and 4=fatal) were recorded. Injury severity was compared using the Fisher's exact test. P values accidents were investigated (n=61 accidents did not lead to any injuries in the occupants, n=7 accidents resulted in minor, and n=6 in severe injuries, and lethal injuries resulted from n=15 accidents). The occupant in the "patient" position was most likely to suffer from deadly injuries (44.9%), followed by the HEMS crew member rear seat (25.0%), compared to lower lethality rates in the other seating positions (9.4-11.2%). Sitting on the HEMS crew member rear seat also was associated with the highest percentage of severe and minor injuries (12.5% each). In HEMS accidents, the patients' position and the HEMS crew member rear seat were found to be at the highest risk for fatal or severe injuries. These results support the urgent requirement of a large international data base for HEMS accidents as a basis for further studies to improve the safety in HEMS missions. Copyright © 2013 Elsevier Ltd. All rights reserved.

Considerations of severe accidents in the design of Korean Next Generation Reactor

International Nuclear Information System (INIS)

Dong Wook Jerng; Choong Sup Byun

1998-01-01

The severe accident is one of the key issues in the design of Korean Next Generation Reactor (KNGR) which is an evolutionary type of pressurized water reactor. As IAEA recommends in TECDOC-801, the design objective of KNGR with regard to safety is provide a sound technical basis by which an imminent off-site emergency response to any circumstance could be practically unnecessary. To implement this design objective, probabilistic safety goals were established and design requirements were developed for systems to mitigate severe accidents. The basic approach of KNGR to address severe accidents is firstly prevent severe accidents by reinforcing its capability to cope with the design basis accidents (DBA) and further with some accidents beyond DBAs caused by multiple failures, and secondly mitigate severe accidents to ensure the retention of radioactive materials in the containment by providing mean to maintain the containment integrity. For severe accident mitigation, KNGR principally takes the concept of ex-vessel corium cooling. To implement this concept, KNGR is equipped with a large cavity and cavity flooding system connected to the in-containment refueling water storage tank. Other major systems incorporated in KNGR are hydrogen igniters and safety depressurization systems. In addition, the KNGR containment is designed to withstand the pressure and temperature conditions expected during the course of severe accidents. In this paper, the design features and status of system designs related with severe accidents will be presented. Also, R and D activities related to severe accident mitigation system design will be briefly described

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

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

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

1993-01-01

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

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

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

Analysis of severe core damage accident progression for the heavy water reactor

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

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

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)

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

Consequences of severe radioactive releases to Nordic Marine environment

Energy Technology Data Exchange (ETDEWEB)

Iosjpe, M. [Norwegian Radiation Protection Authority (NRPA) (Norway); Isaksson, M. [Univ. of Gothenburg (Sweden); Joensen, H.P. [Froskaparsetur Foeroya. Faroe Islands, Torshavn (Denmark); Lahtinen, J. [Radiation and Nuclear Safety Authority (Finland); Logemann, K. [Univ. of Iceland (Iceland); Palsson, S.E. [Geislavarnir Rikisins (Iceland); Roos, P. [Technical Univ. of Denmark. Risoe DTU, Roskilde (Denmark); Suolanen, V. [Technical Research Centre of Finland (Finland)

2013-02-15

In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW{sub th} reactor size were used. Based on source term analyses, the

Consequences of severe radioactive releases to Nordic Marine environment

International Nuclear Information System (INIS)

Iosjpe, M.; Isaksson, M.; Joensen, H.P.; Lahtinen, J.; Logemann, K.; Palsson, S.E.; Roos, P.; Suolanen, V.

2013-02-01

In the report, consequences of hypothetical severe nuclear accidents releases to Nordic marine environment are preliminary considered. The considered marine area comprises the Baltic Sea (Sweden, Denmark, Finland) and the North Atlantic (Iceland, Faroes, Norway) areas. The hypothetical severe nuclear accidents can be related to nuclear power plants, nuclear powered submarines or ice-breakers. Quite comprehensive survey on radioactive source terms of extremely severe nuclear power and submarine accidents has been done. This enables to estimate more realistically possible radioactive releases of various elements and nuclides to marine environment. One recent reference is of course the Fukushima accident and estimated releases there. The marine flows and dilution circumstances around the Nordic nuclear power plants and in the Baltic Sea area in general, has been studied. Respectively marine flows related to Iceland and Faroes coasts are considered with measured data and with preliminary 3D-model simulations. The substantial depth of sea water in the North Atlantic affect vertical concentration profiles to some extent. At Icelandic or Faroese waters, a potential submarine accident would likely occur in a well defined water mass, and radioactivity from the accident would be detected and spread with the flow regime of the water mass in the world ocean. Based on hypothetical severe accidents scenarios, preliminary consequence calculations has been done. It should be emphasised that the considered severe accident cases, considered in this study, do not directly attach any specific Nordic nuclear power plant or any specific submarine type. The considered radioactive releases will, however, provide specified references for more extensive consideration of environmental consequences of severe - or minor - radioactive releases to Nordic marine environment. As a reference, the release amounts from a 3000 MW th reactor size were used. Based on source term analyses, the chosen

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

International Nuclear Information System (INIS)

Ahn, K. I.; Kim, D. H.

2003-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2003-10-01

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

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

Perspectives on phenomenology and simulation of severe accident in light water reactors

International Nuclear Information System (INIS)

Sugimoto, Jun

2014-01-01

Severe accident phenomena in light water reactors (LWRs) are generally characterized by their physically and chemically complex processes involved with high temperature core melt, multi-component and multi-phase flows, transport of radioactive materials and sometimes highly non-equilibrium state. Severe accident phenomenology is usually categorized into four phases; (1) fuel degradation, (2) in-vessel phenomena, (3) ex-vessel phenomena and (4) fission product release and transport. Among these, ex-vessel phenomena consist of five subcategories; 1) direct containment heating, 2) fuel coolant interaction (steam explosion), 3) molten core concrete interaction, 4) hydrogen behaviour and control and 5) containment failure/leakage. In the field of simulation of severe accident, severe accident analytical codes have been developed in the United States, EU and Japan, such as MAAP, MELCOR, ASTEC, THALES and SAMPSON. Many different kinds of analytical codes for the specific severe accident phenomena have also been developed worldwide. After the accident at Fukushima Daiichi Nuclear Power Station, review of severe accident research issues has been conducted and several issues are reconsidered, such as effects of BWR core degradation behaviors, sea water injection, pool scrubbing under rapid depressurization, containment failure/leakage and re-criticality. Some new experimental and analytical efforts have been started after the Fukushima accident. The present paper describes the perspectives on phenomenology and simulation of severe accident in LWRs, with the emphasis of insights obtained in the review of Fukushima accident. (author)

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

International Nuclear Information System (INIS)

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

2005-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-03-15

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

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

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)

Development of instrumentation systems for severe accidents. 4. New accident tolerant in-containment pressure transducer for containment pressure monitoring system

International Nuclear Information System (INIS)

Oba, Masato; Teruya, Kuniyuki; Yoshitsugu, Makoto; Ikeuchi, Takeshi

2015-01-01

The accident at Tokyo Electric Power Company's Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty in measuring important parameters for monitoring plant conditions. Therefore, we have studied the TF-1 accident to select the important parameters that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems that could measure the parameters in severe accident conditions. Mitsubishi Heavy Industries, LTD (MHI) developed a new accident tolerant containment pressure monitoring system and demonstrated that the monitoring system could endure extremely harsh environmental conditions that envelop severe accident environmental conditions inside a containment such as maximum operating temperature of up to 300degC and total integrated dose (TID) of 1 MGy gamma. The new containment pressure monitoring system comprises of a strain gage type pressure transducer and a mineral insulated (MI) cable with ceramic connectors, which are located in the containment, and a strain measuring amplifier located outside the containment. Less thermal and radiation degradation is achieved because of minimizing use of organic materials for in-containment equipment such as the transducer and connectors. Several tests were performed to demonstrate the performance and capability of the in-containment equipment under severe accident environmental conditions and the major steps in this testing were run in the following test sequences: (1) the baseline functional tests (e.g., repeatability, non-linearity, hysteresis, and so on) under normal conditions, (2) accident radiation testing, (3) seismic testing, and (4) steam/temperature test exposed to simulated severe accident environmental conditions. The test results demonstrate that the new pressure transducer can endure the simulated severe accident conditions. (author)

Severe fuel damage projects

International Nuclear Information System (INIS)

Sdouz, G.

1987-10-01

After the descriptions of the generation of a Severe Fuel Damage Accident in a LWR the hypothetical course of such an accident is explained. Then the most significant projects are described. At each project the experimental facility, the most important results and the concluding models and codes are discussed. The selection of the projects is concentrated on the German Projekt Nukleare Sicherheit (PNS), tests performed at the Idaho National Engineering Laboratory (INEL) and smaller projects in France and Great Britain. 25 refs., 26 figs. (Author)

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

International Nuclear Information System (INIS)

Frid, W.

2002-01-01

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

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

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.

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

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)

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)

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

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

High-speed radiation dose calculations for severe accidents using INDOS

International Nuclear Information System (INIS)

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

1992-01-01

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

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

Analysis of hot leg natural circulation under station blackout severe accident

International Nuclear Information System (INIS)

Deng Jian; Cao Xuewu

2007-01-01

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

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

International Nuclear Information System (INIS)

Hashimoto, Kazuichiro

2000-11-01

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

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

SARNET. Severe Accident Research Network - key issues in the area of source term

International Nuclear Information System (INIS)

Giordano, P.; Micaelli, J.C.; Haste, T.; Herranz, L.

2005-01-01

About fifty European organisations integrate in SARNET (Network of Excellence of the EU 6 th Framework Programme) their research capacities in resolve better the most important remaining uncertainties and safety issues concerning existing and future Nuclear Power Plants (NPPs) under hypothetical Severe Accident (SA) conditions. Wishing to maintain a long-lasting cooperation, they conduct three types of activities: integrating activities, spreading of excellence and jointly executed research. This paper summarises the main results obtained by the network after the first year, giving more prominence to those from jointly executed research in the Source Term area. Integrating activities have been performed through different means: the ASTEC integral computer code for severe accident transient modelling, through development of PSA2 methodologies, through the setting of a structure for definition of evolving R and D priorities and through the development of a web-network of data bases that hosts experimental data. Such activities have been facilitated by the development of an Advanced Communication Tool. Concerning spreading of excellence, educational courses covering Severe Accident Analysis Methodology and Level 2 PSA have been set up, to be given in early 2006. A detailed text book on Severe Accident Phenomenology has been designed and agreed amongst SARNET members. A mobility programme for students and young researchers is being developed, some detachments are already completed or in progress, and examples are quoted. Jointly executed research activities concern key issues grouped in the Corium, Containment and Source Term areas. In Source Term, behaviour of the highly radio-toxic ruthenium under oxidising conditions (like air ingress) for HBU and MOX fuel has been investigated. First modelling proposals for ASTEC have been made for oxidation of fuel and of ruthenium. Experiments on transport of highly volatile oxide ruthenium species have been performed. Reactor

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

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

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.

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

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

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

2014-01-01

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

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

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

1991-01-01

A methodology has been developed which allows the accident probabilities associated with one accident-severity category scheme to be transferred to another severity category scheme. The methodology requires that the schemes use a common set of parameters to define the categories. The transfer of accident probabilities is based on the relationships between probability of occurrence and each of the parameters used to define the categories. Because of the lack of historical data describing accident environments in engineering terms, these relationships may be difficult to obtain directly for some parameters. Numerical models or experienced judgement are often needed to obtain the relationships. These relationships, even if they are not exact, allow the accident probability associated with any severity category to be distributed within that category in a manner consistent with accident experience, which in turn will allow the accident probability to be appropriately transferred to a different category scheme

Contribution of prototypic material tests on the Plinius platform to the study of nuclear reactor severe accident; Contribution des essais en materiaux prototypiques sur la plate-forme Plinius a l'etude des accidents graves de reacteurs nucleaires

Energy Technology Data Exchange (ETDEWEB)

Journeau, Ch

2008-01-15

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

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

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

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

Energy Technology Data Exchange (ETDEWEB)

Hashimoto, Kazuichiro [ed.

2000-11-01

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

Human error as the root cause of severe accidents at nuclear reactors

International Nuclear Information System (INIS)

Kovács Zoltán; Rýdzi, Stanislav

2017-01-01

A root cause is a factor inducing an undesirable event. It is feasible for root causes to be eliminated through technological process improvements. Human error was the root cause of all severe accidents at nuclear power plants. The TMI accident was caused by a series of human errors. The Chernobyl disaster occurred after a badly performed test of the turbogenerator at a reactor with design deficiencies, and in addition, the operators ignored the safety principles and disabled the safety systems. At Fukushima the tsunami risk was underestimated and the project failed to consider the specific issues of the site. The paper describes the severe accidents and points out the human errors that caused them. Also, provisions that might have eliminated those severe accidents are suggested. The fact that each severe accident occurred on a different type of reactor is relevant – no severe accident ever occurred twice at the same reactor type. The lessons learnt from the severe accidents and the safety measures implemented on reactor units all over the world seem to be effective. (orig.)

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

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

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.

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

[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.

Mitigation of Hydrogen Hazards in Severe Accidents in Nuclear Power Plants

International Nuclear Information System (INIS)

2011-07-01

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

Review of current status for designing severe accident management support system

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwang Sub

2000-05-01

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

Review of current status for designing severe accident management support system

International Nuclear Information System (INIS)

Jeong, Kwang Sub

2000-05-01

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

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

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

Analysis of initial events following hypothetical criticality of a transport flask

International Nuclear Information System (INIS)

Barbry, F.; Bonhomme, C.; Brown, M.L.; Hague, P.; Mather, D.J.; Shaw, P.M.

1984-01-01

This report deals with the estimation of possible consequences, eg energy release, temperatures reached etc, of such a hypothetical accident in a particular notional transport package design. This particular study examines the situation if criticality occurs during unloading or refilling of a PWR flask. In the first instance, an idealised model has been chosen in order to develop the calculational techniques; it is not initself a realistic accident representation

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

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.

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-01

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

Analysis of Maximum Reasonably Foreseeable Accidents for the Yucca Mountain Draft Environmental Impact Statement (DEIS)

International Nuclear Information System (INIS)

Ross, S.B.; Best, R.E.; Maheras, S.J.; McSweeney, T.I.

2001-01-01

Accidents could occur during the transportation of spent nuclear fuel and high-level radioactive waste. This paper describes the risks and consequences to the public from accidents that are highly unlikely but that could have severe consequences. The impact of these accidents would include those to a collective population and to hypothetical maximally exposed individuals (MEIs). This document discusses accidents with conditions that have a chance of occurring more often than 1 in 10 million times in a year, called ''maximum reasonably foreseeable accidents''. Accidents and conditions less likely than this are not considered to be reasonably foreseeable

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)

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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)

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

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

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

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

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)

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

International Nuclear Information System (INIS)

Chang, Hyun Sop

1996-02-01

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

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

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.

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

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