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Sample records for bwr severe accident

  1. 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.

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Chemical factors affecting fission product transport in BWR severe accidents

    International Nuclear Information System (INIS)

    Chemical changes may significantly alter physical properties of fission product materials, and hence their state and transport rate. Thus, it is possible that an appropriate accounting of chemical change could have a large impact on transport model results. This paper will describe how the chemical reactions of Cs, I, and Te are being implemented in the transport model that is used in the Severe Accident Sequence Analysis (SASA) Program at Oak Ridge National Laboratory (ORNL)

  6. Severe accident instrumentation systems for BWR water level and temperature in primary containment vessel measurements

    International Nuclear Information System (INIS)

    The severe accident at TEPCO's Fukushima Daiichi nuclear power station (TF1 accident) in March 2011 brought the lost of the functions of many instrumentation systems. In order to enable the measurements of the important parameters such as reactor water level, temperature and so on even in a case such as the TF1 accident occurs, severe accident instrumentation systems are being developed. In this paper, new system configurations of BWR water level measurement and temperature measurement in primary containment vessels are proposed. Then performance tests for prototype sensors of these measurement systems under high temperature conditions are described. (author)

  7. Role of BWR secondary containments in severe accident mitigation: issues and insights from recent analyses

    International Nuclear Information System (INIS)

    All commercial boiling water reactor (BWR) plants in the US employ primary containments of the pressure suppression design. These primary containments are surrounded and enclosed by a secondary containment consisting of a reactor building and refueling bay (MK I and MK II designs), a shield building, auxiliary building and fuel building (MK III), or an auxiliary building and enclosure building (Grand Gulf style MK III). Although secondary containment designs are highly plant specific, their purpose is to minimize the ground level release of radioactive material for a spectrum of traditional design basis accidents. While not designed for severe accident mitigation, these secondary containments might also reduce the radiological consequences of severe accidents. This issue is receiving increasing attention due to concerns that BWR MK I primary containment integrity would be lost should a significant mass of molten debris escape the reactor vessel during a severe accident. This paper presents a brief overview of domestic BWR secondary containment designs and highlights plant-specific features that could influence secondary containment severe accident survivability and accident mitigation effectiveness. Current issues surrounding secondary containment performance are discussed, and insights gained from recent ORNL secondary containment studies of Browns Ferry, Peach Bottom, and Shoreham are presented. Areas of significant uncertainty are identified and recommendations for future research are presented

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

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

    International Nuclear Information System (INIS)

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

  10. Role of BWR MK I secondary containments in severe accident mitigation

    International Nuclear Information System (INIS)

    The recent advent of detailed containment analysis codes such as CONTAIN and MELCOR has facilitated the development of the first large-scale, architectural-based BWR secondary containment models. During the past year ORNL has developed detailed, plant-specific models of the Browns Ferry and Peach Bottom secondary containments, and applied these models in a variety of studies designed to evaluate the role and effectiveness of BWR secondary containments in severe accident mitigation. The topology and basis for these models is discussed, together with some of the emerging insights from these studies

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

    International Nuclear Information System (INIS)

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

  12. Assessment of severe accident prevention and mitigation features: BWR, Mark III containment design

    International Nuclear Information System (INIS)

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

  13. A direct comparison of MELCOR 1.8.3 and MAAP4 results for several PWR ampersand BWR accident sequences

    International Nuclear Information System (INIS)

    This paper presents a comparison of calculations of severe accident progression for several postulated accident sequences for representative Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) nuclear power plants performed with the MELCOR 1.8.3 and the MAAP4 computer codes. The PWR system examined in this study is a 1100 MWe system similar in design to a Westinghouse 3-loop plant with a large dry containment; the BWR is a 1100 MWe system similar in design to General Electric BWR/4 with a Mark I containment. A total of nine accident sequences were studied with both codes. Results of these calculations are compared to identify major differences in the timing of key events in the calculated accident progression or other important aspects of severe accident behavior, and to identify specific sources of the observed differences

  14. Effects of improved modeling on best estimate BWR severe accident analysis

    International Nuclear Information System (INIS)

    Since 1981, ORNL has completed best estimate studies analyzing several dominant BWR accident scenarios. These scenarios were identified by early Probabilistic Risk Assessment (PRA) studies and detailed ORNL analysis complements such studies. In performing these studies, ORNL has used the MARCH code extensively. ORNL investigators have identified several deficiencies in early versions of MARCH with regard to BWR modeling. Some of these deficiencies appear to have been remedied by the most recent release of the code. It is the purpose of this paper to identify several of these deficiencies. All the information presented concerns the degraded core thermal/hydraulic analysis associated with each of the ORNL studies. This includes calculations of the containment response. The period of interest is from the time of permanent core uncovery to the end of the transient. Specific objectives include the determination of the extent of core damage and timing of major events (i.e., onset of Zr/H2O reaction, initial clad/fuel melting, loss of control blade structure, etc.). As mentioned previously the major analysis tool used thus far was derived from an early version of MARCH. BWRs have unique features which must be modeled for best estimate severe accident analysis. ORNL has developed and incorporated into its version of MARCH several improved models. These include (1) channel boxes and control blades, (2) SRV actuations, (3) vessel water level, (4) multi-node analysis of in-vessel water inventory, (5) comprehensive hydrogen and water properties package, (6) first order correction to the ideal gas law, and (7) separation of fuel and cladding. Ongoing and future modeling efforts are required. These include (1) detailed modeling for the pressure suppression pool, (2) incorporation of B4C/steam reaction models, (3) phenomenological model of corium mass transport, and (4) advanced corium/concrete interaction modeling. 10 references, 17 figures, 1 table

  15. Quantification of severe accidents source terms of BWR 4 reactor with Mark I containment using source term code package

    International Nuclear Information System (INIS)

    Severe accident source terms of a nuclear power plant which employs a BWR4 reactor with a Mark I containment are quantified with the Source Term Code Package (STCP). Accident scenarios selected for source terms analyses are defined based on the Probabilistic Risk Assessment (PRA) results of accident sequence grouping, containment responses, containment phenomenological event trees, and release category analyses of studies. Included in the paper is a brief description of the structure and major features of STCP together with the modifications made to the code package for the present analysis, the plant model adopted for the STCP source terms quantifications; a presentation and discussion of the source terms as predicted by the STCP for the ten accident sequences analyzed. (orig.)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  17. A synergistic use of CFD, experiments and effective convectivity model to reduce uncertainty in BWR severe accident analysis

    International Nuclear Information System (INIS)

    In a previous work we presented an analysis approach developed to effectively and accurately assess thermal loads on vessel and structures in a Boiling Water Reactor (BWR) lower head during a severe accident. Central to the assessment is the Effective Convectivity Model (ECM) that makes use of experimental heat transfer correlations to capture the effect of turbulent natural convection in a volumetrically heated liquid pool, while retaining the pool three-dimensional energy splitting and ability to represent local heat transfer effects. Thanking to its features, the ECM is unique in enabling calculations of complex heat transfer phenomena during long severe accident transients that would not be otherwise feasible using higher-fidelity methods such as Computational Fluid Dynamics (CFD). Efficiency notwithstanding, the natural questions are: (i) how good are those ECM-calculated results, and, (ii) if required, what can be done (with the highest return-on-investment) to improve the quality of ECM prediction results. The approach refers to experiments and CFD simulations as the main resources to address (i) and (ii). However, validation of ECM against simulant-fluid experiments by itself does not reveal deficiencies (due to non-prototypicality factors). In the present work we focus on the use of CFD-based numerical 'experiments' to identify and quantify source of epistemic uncertainty in the calculated thermal loads due to modeling assumptions in ECM. Specifically, heat transfer correlations that underlie the ECM are obtained as surface-averaged (even though implemented as spatially distributed) and derived from experiments conducted at different geometries and using fluids that are not reactor prototypical (molten corium in the present case of severe accident). The CFD simulations exhibit so-called fluid Prandtl number effect on local peaking of the pool's downward heat flux for corium as working fluid. The main premise is a synergistic use of a fast-running model

  18. Severe accident analyses of a BWR with MAAP5 code. Station blackout and large-break LOCA

    International Nuclear Information System (INIS)

    Calculations were performed for a station blackout (TBU) sequence and a large-break loss-of-coolant accident (AE) sequence of a typical BWR-5 plant with modified Mark-II type containment by the MAAP5 code. The core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment including hydrogen production were investigated. Sensitivity analyses focusing on direct contact heating (DCH) and zirconium oxidation, which affect on the consequences of severe accidents, were also performed. If extensive DCH does not occur in the TBU sequence, failure of the containment vessel can be postponed. On the other hand, concrete ablation at a floor and a side wall in the pedestal due to molten core - concrete interaction (MCCI) significantly increases, because a large amount of debris with high temperature stays inside the pedestal. Although the hydrogen production is affected by the zirconium oxidation model, the differences of hydrogen production are within ± 10% in the case of TBU sequence. (author)

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

    International Nuclear Information System (INIS)

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

  20. Assessment of two BWR accident management strategies

    International Nuclear Information System (INIS)

    Candidate mitigative strategies for the management of in-vessel events during the late phase (after-core degradation has occurred) of postulated boiling water reactor (BWR) severe accidents were considered at Oak Ridge National Laboratory (ORNL) during 1990. The identification of new strategies was subject to the constraint that they should, to the maximum extent possible, make use of the existing equipment and water resources of the BWR facilities, and not require major equipment modifications or additions. As a result of this effort, two of these candidate strategies were recommended for further assessment. The first was a strategy for containment flooding to maintain the core and structural debris within the reactor vessel in the event that vessel injection cannot be restored to terminate a severe accident sequence. The second strategy pertained to the opposite case, for which vessel injection would be restored after control blade melting had begun; its purpose was to provide an injection source of borated water at the concentration necessary to preclude criticality upon recovering a damaged BWR core. Assessments of these two strategies were performed during 1991 and this paper provides a discussion of the motivation for and purpose of these strategies, and the potential for their success. ((orig.))

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

    International Nuclear Information System (INIS)

    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)

  2. Water level measurement system in reactor pressure vessel of BWR and hydrogen concentration monitoring system for severe accident

    International Nuclear Information System (INIS)

    TEPCO's Fukushima Daiichi Nuclear Power Station Accident caused severe accident to lose functions of many instrumentation systems. As a result, many important plant parameters couldn't be monitored. In order to monitor plant parameters in the case of severe accident, new instrumentation systems available in the severe conditions are being developed. Water level in reactor pressure vessel and hydrogen concentration in primary containment vessel are one of the most important parameters. Performance test results about water level measurement sensor and hydrogen sensor in severe environmental conditions are described. (author)

  3. Contain calculations of debris conditions adjacent to the BWR Mark I drywell shell during the later phases of a severe accident

    International Nuclear Information System (INIS)

    Best estimate CONTAIN calculations have recently been performed by the BWR Severe Accident Technology (BWRSAT) Program at Oak Ridge National Laboratory to predict the primary containment response during the later phases of an unmitigated low-pressure Short Term Station Blackout at the Peach Bottom Atomic Power Station. Debris pour conditions leaving the failed reactor vessel are taken from the results of best estimate BWRSAR analyses that are based upon an assumed metallic debris melting temperature of 2750/degree/F (1783 K) and an oxide debris melting temperature of 4350/degree/F (2672 K). Results of the CONTAIN analysis for the case without sprays indicate failure of the drywell seals due to the extremely hot atmospheric conditions extant in the drywell. The maximum calculated temperature of the debris adjacent to the drywell shell is less than the melting temperature of the shell, yet the sustained temperatures may be sufficient to induce primary containment pressure boundary failure by the mechanism of creep-rupture. It is also predicted that a significant portion of the reactor pedestal wall is ablated during the period of the calculation. Nevertheless, the calculated results are recognized to be influenced by large modeling uncertainties. Several deficiencies in the application of the CORCON module within the CONTAIN code to BWR severe accident sequences are identified and discussed. 5 refs., 9 figs., 4 tabs.,

  4. Application of Bayesian nonparametric models to the uncertainty and sensitivity analysis of source term in a BWR severe accident

    International Nuclear Information System (INIS)

    A full-scope method is constructed to reveal source term uncertainties and to identify influential inputs during a severe accident at a nuclear power plant (NPP). An integrated severe accident code, MELCOR Ver. 1.8.5, is used as a tool to simulate the accident similar to that occurred at Unit 2 of the Fukushima Daiichi NPP. In order to figure out how much radioactive materials are released from the containment to the environment during the accident, Monte Carlo based uncertainty analysis is performed. Generally, in order to evaluate the influence of uncertain inputs on the output, a large number of code runs are required in the global sensitivity analysis. To avoid the laborious computational cost for the global sensitivity analysis via MELCOR, a surrogate stochastic model is built using a Bayesian nonparametric approach, Dirichlet process. Probability distributions derived from uncertainty analysis using MELCOR and the stochastic model show good agreement. The appropriateness of the stochastic model is cross-validated through the comparison with MELCOR results. The importance measure of uncertain input variables are calculated according to their influences on the uncertainty distribution as first-order effect and total effect. The validity of the present methodology is demonstrated through an example with three uncertain input variables. - Highlights: • A method of source term uncertainty and sensitivity analysis is proposed. • Source term in Fukushima Daiichi NPP severe accident is demonstrated. • Uncertainty distributions of source terms show non-standard shapes. • A surrogate model for integrated code is constructed by using Dirichlet process. • Importance ranking of influential input variables is obtained

  5. Performance of the primary containment of a BWR during a severe accident whit the code RELAP/SCDAPSIM; Comportamiento del contenedor primario de un reactor BWR durante un accidente severo con el codigo RELAP/SCDAPSIM

    Energy Technology Data Exchange (ETDEWEB)

    Castillo G, F.

    2015-07-01

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

  6. Severe accident phenomena

    International Nuclear Information System (INIS)

    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.)

  7. 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.

  8. SEVERE ACCIDENT MANAGEMENT TRAINING

    International Nuclear Information System (INIS)

    The purpose of this paper is (a) to define the International Atomic Energy Agency's role in the area of severe accident management training, (b) to briefly describe the status of representative severe accident analysis tools designed to support development and validation of accident management guidelines, and more recently, simulate the accident with sufficient accuracy to support the training of technical support and reactor operator staff, and (c) provide an overview of representative design-specific accident management guidelines and training. Since accident management and the development of accident management validation and training software is a rapidly evolving area, this paper is also intended to evolve as accident management guidelines and training programs are developed to meet different reactor design requirements and individual national requirements

  9. Severe accident recriticality analyses (SARA)

    DEFF Research Database (Denmark)

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

    2001-01-01

    Recriticality in a BWR during reflooding of an overheated partly degraded core, i.e. with relocated control rods, has been studied for a total loss of electric power accident scenario. In order to assess the impact of recriticality on reactor safety, including accident management strategies, the ...

  10. Code comparison with MAAP 3.0 and March 3 (-STCP) for Nordic BWR and PWR plants to evaluate uncertainties in severe accident phenomena

    International Nuclear Information System (INIS)

    This study has been carried out within the framework of the Nordic NKA-AKTI-130-project whose participants are from Denmark, Finland and Sweden. The study is financed partly by the Nordic liaison committee for atomic energy and partly by national organisations. The goals of the study have been to achieve a common Nordic understanding of the capabilities of the severe accident codes MAAP 3.0 /1, 2/ and March 3-STCP /3/ and to evaluate uncertainties in severe accident phenomena by performing benchmark calculations and related sensitivity analyses for the existing Nordic power plants. The MAAP 3.0 code, which is an integrated thermal hydraulic and aerosol code, has been the main analysis tool in severe accident analyses in Sweden and Finland. Danish organisations have used the Source Term Code Package system (Mod 1.0) which is composed of several separate codes such as March 3, TRAPMELT etc. When plant specific design features are analyzed, a sensitivity type of study with a code system like MAAP 3.0 is an efficient tool. Experimental data for validation of code systems modelling the complex phenomena involved in severe accidents are, however, limited. It is in this situation valuable to compare models and results for two code systems developed by different organizations

  11. Review of the status of validation of the computer codes used in the severe accident source term reassessment study (BMI-2104). [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    Kress, T. S. [comp.

    1985-04-01

    The determination of severe accident source terms must, by necessity it seems, rely heavily on the use of complex computer codes. Source term acceptability, therefore, rests on the assessed validity of such codes. Consequently, one element of NRC's recent efforts to reassess LWR severe accident source terms is to provide a review of the status of validation of the computer codes used in the reassessment. The results of this review is the subject of this document. The separate review documents compiled in this report were used as a resource along with the results of the BMI-2104 study by BCL and the QUEST study by SNL to arrive at a more-or-less independent appraisal of the status of source term modeling at this time.

  12. Twenty-fifth water reactor safety information meeting: Proceedings. Volume 1: Plenary sessions; Pressure vessel research; BWR strainer blockage and other generic safety issues; Environmentally assisted degradation of LWR components; Update on severe accident code improvements and applications

    International Nuclear Information System (INIS)

    This three-volume report contains papers presented at the conference. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from France, Japan, Norway, and Russia. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting. This volume contains the following information: (1) plenary sessions; (2) pressure vessel research; (3) BWR strainer blockage and other generic safety issues; (4) environmentally assisted degradation of LWR components; and (5) update on severe accident code improvements and applications. Selected papers have been indexed separately for inclusion in the Energy Science and Technology Database

  13. Management of severe accidents

    International Nuclear Information System (INIS)

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

  14. Severe accident analysis of a representative LWR plant with MAAP5.01 and MELCOR2.1. Comparison of station blackout analysis for a BWR-5/advanced Mark-II containment type plant

    International Nuclear Information System (INIS)

    In order to investigate the differences in characteristics of MAAP5.01 and MELCOR2.1, which are dedicated codes to evaluate severe accident progression, severe accident analyses for a TBU sequence (station blackout with no emergency power supply and no recovery of short- and long-term A/C power) in a BWR-5/advanced Mark-II containment type plant were conducted by using the two codes. Based on the analysis that input settings of the decay heat of fuel, the failure criteria of fuel cladding and core support plate, and zirconium – water interaction model are adjusted between these codes, the hydrodynamic response inside the reactor pressure vessel (RPV) in the early phase (until the onset of fuel relocation) of the TBU sequence is shown to be in good agreement. However, significant differences are observed in the onset timing of the major physical phenomena after the core support plate failure. It is inferred that these disagreements are primarily caused by the differences in characteristics of analytical models in each code, such as debris relocation, coolant channel blockage, entrainment and quench of the molten debris jet in water pool, failure of the RPV lower head, and molten core – concrete interaction (MCCI). (author)

  15. Severe accident testing of electrical penetration assemblies

    International Nuclear Information System (INIS)

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

  16. Severe accident testing of electrical penetration assemblies

    Energy Technology Data Exchange (ETDEWEB)

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

    1989-11-01

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

  17. Serious accidents on boiling water reactors (BWR)

    International Nuclear Information System (INIS)

    This short document describes, first, the specificities of boiling water reactors (BWRs) with respect to PWRs in front of the progress of a serious accident, and then, the strategies of accident management: restoration of core cooling, water injection, core flooding, management of hydrogen release, depressurization of the primary coolant circuit, containment spraying, controlled venting, external vessel cooling, erosion of the lower foundation raft by the corium). (J.S.)

  18. Severe accident management program at Cofrentes Nuclear Power Plant

    International Nuclear Information System (INIS)

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

  19. Severe Accident Recriticality Analyses (SARA)

    International Nuclear Information System (INIS)

    Recriticality in a BWR has been studied for a total loss of electric power accident scenario. In a BWR, the B4C control rods would melt and relocate from the core before the fuel during core uncovery and heat-up. If electric power returns during this time-window unborated water from ECCS systems will start to reflood the partly control rod free core. Recriticality might take place for which the only mitigating mechanisms are the Doppler effect and void formation. In order to assess the impact of recriticality on reactor safety, including accident management measures, the following issues have been investigated in the SARA project: 1. the energy deposition in the fuel during super-prompt power burst, 2. the quasi steady-state reactor power following the initial power burst and 3. containment response to elevated quasi steady-state reactor power. The approach was to use three computer codes and to further develop and adapt them for the task. The codes were SIMULATE-3K, APROS and RECRIT. Recriticality analyses were carried out for a number of selected reflooding transients for the Oskarshamn 3 plant in Sweden with SIMULATE-3K and for the Olkiluoto 1 plant in Finland with all three codes. The core state initial and boundary conditions prior to recriticality have been studied with the severe accident codes SCDAP/RELAP5, MELCOR and MAAP4. The results of the analyses show that all three codes predict recriticality - both superprompt power bursts and quasi steady-state power generation - for the studied range of parameters, i. e. with core uncovery and heat-up to maximum core temperatures around 1800 K and water flow rates of 45 kg/s to 2000 kg/s injected into the downcomer. Since the recriticality takes place in a small fraction of the core the power densities are high which results in large energy deposition in the fuel during power burst in some accident scenarios. The highest value, 418 cal/g, was obtained with SIMULATE-3K for an Oskarshamn 3 case with reflooding rate

  20. Severe Accident Recriticality Analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

    Frid, W. [Swedish Nuclear Power Inspectorate, Stockholm (Sweden); Hoejerup, F. [Risoe National Lab. (Denmark); Lindholm, I.; Miettinen, J.; Puska, E.K. [VTT Energy, Helsinki (Finland); Nilsson, Lars [Studsvik Eco and Safety AB, Nykoeping (Sweden); Sjoevall, H. [Teoliisuuden Voima Oy (Finland)

    1999-11-01

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

  1. Influence of boron vapor on transport behavior of deposited CsI during heating test simulating a BWR severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Sato, Isamu, E-mail: sato.isamu@jaea.go.jp; Onishi, Takashi; Tanaka, Kosuke; Iwasaki, Maho; Koyama, Shin-ichi

    2015-06-15

    In order to evaluate influence of B on the release and transport of Cs and I during severe accidents, basic experiments have been performed on the interaction between deposited Cs/I compounds and vapor/aerosol B compounds. CsI and B{sub 2}O{sub 3} were utilized as a Cs/I compound and a B compound, respectively. Deposited CsI on the thermal gradient tube (TGT) at temperatures ranging from 423 K to 1023 K was reacted with vapor/aerosol B{sub 2}O{sub 3}, and then observed how it changed Cs/I deposition profiles. As a result, vapor/aerosol B{sub 2}O{sub 3} stripped a portion of deposited CsI within a temperature range from 830 K to 920 K to make gaseous CsBO{sub 2} and I{sub 2}. In addition, gaseous I{sub 2} was re-deposited at a temperature range from 530 K to 740 K, while CsBO{sub 2} travelled through the sampling tubes and filters without deposition. It is evident that B enables Cs compounds such as CsBO{sub 2} to transport Cs to the colder regions.

  2. Severe accident insights from the Brunswick IPE

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. 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.

  6. Identification and initial assessment of candidate BWR late-phase in-vessel accident management strategies

    Energy Technology Data Exchange (ETDEWEB)

    Hodge, S.A.

    1991-04-15

    Work sponsored by the United States Nuclear Regulatory Commission (USNRC) to identify and perform preliminary assessments of candidate BWR (boiling water reactor) in-vessel accident management strategies was completed at Oak Ridge National Laboratory (ORNL) during fiscal year 1990. Mitigative strategies for containment events have been the subject of a companion study at Brookhaven National Laboratory. The focus of this Oak Ridge effort was the development of new strategies for mitigation of the late phase events, that is, the events that would occur in-vessel after the onset of significant core damage. The work began with an investigation of the current status of BWR in-vessel accident management procedures and proceeded through a preliminary evaluation of several candidate new strategies. The steps leading to the identification of the candidate strategies are described. The four new candidate late-phase (in-vessel) accident mitigation strategies identified by this study and discussed in the report are: (1) keep the reactor vessel depressurized; (2) restore injection in a controlled manner; (3) inject boron if control blade damage has occurred; and (4) containment flooding to maintain core and structural debris in-vessel. Additional assessments of these strategies are proposed.

  7. Severe accidents, a US approach

    International Nuclear Information System (INIS)

    The attitude of the American nuclear industry and the regulatory authorities in the United States toward severe accidents has often seemed ambivalent. It was common a few years ago to assume the position that severe accidents should not be included in the design basis of the plant. This view was associated with the concept of the maximum credible accident. A severe accident that would lead to a large release of fission products from the reactor core was simply regarded as having so low a likelihood as not to be credible. That does not mean that it had a zero probability of occurring. Because of the way the plant was designed, built, and operated, severe accidents were regarded as having a low enough probability that no further special measures were necessary regarding them. (author)

  8. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

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

  9. Analysis of the FeCrAl Accident Tolerant Fuel Concept Benefits during BWR Station Blackout Accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

    Iron-chromium-aluminum (FeCrAl) alloys are being considered for fuel concepts with enhanced accident tolerance. FeCrAl alloys have very slow oxidation kinetics and good strength at high temperatures. FeCrAl could be used for fuel cladding in light water reactors and/or as channel box material in boiling water reactors (BWRs). To estimate the potential safety gains afforded by the FeCrAl concept, the MELCOR code was used to analyze a range of postulated station blackout severe accident scenarios in a BWR/4 reactor employing FeCrAl. The simulations utilize the most recently known thermophysical properties and oxidation kinetics for FeCrAl. Overall, when compared to the traditional Zircaloy-based cladding and channel box, the FeCrAl concept provides a few extra hours of time for operators to take mitigating actions and/or for evacuations to take place. A coolable core geometry is retained longer, enhancing the ability to stabilize an accident. Finally, due to the slower oxidation kinetics, substantially less hydrogen is generated, and the generation is delayed in time. This decreases the amount of non-condensable gases in containment and the potential for deflagrations to inhibit the accident response.

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

    International Nuclear Information System (INIS)

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

  11. Severe accident management guidelines tool

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. The vver severe accident management

    International Nuclear Information System (INIS)

    The basic approach to the VVER safety management is based on the defence-in-depth principle the main idea of which is the multiplicity of physical barriers on the way of dangerous propagation on the one hand and the diversity of measures to protect each of them on the other hand. The main events of severe accident with loss of core cooling at NPP with WWER can be represented as a sequence of NPP states, in which each subsequent state is more severe than the previous one. The following sequence of states of the accident progression is supposed to be realistic and the most probable: -) loss of efficient core cooling; -) core melting, relocation of the molten core to the lower head and molten pool formation, -) reactor vessel damage, and -) containment damage and fission products release. The objectives of accident management at the design basis stage, the determining factors and appropriate determining parameters of processes are formulated in this paper. The same approach is used for the estimation of processes parameters at beyond design basis accident progression. The accident management goals and the determining factors and parameters are also listed in that case which is characterized by the loss of integrity of the fuel cladding. The accident management goal at the stage of core melt relocation implies the need for an efficient core-catcher

  14. Porosity effects during a severe accident

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-09-15

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

  15. A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

    International Nuclear Information System (INIS)

    The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

  16. Severe accident source term reassessment

    International Nuclear Information System (INIS)

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

  17. Safety against releases in severe accidents. Final report

    International Nuclear Information System (INIS)

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

  18. Safety against releases in severe accidents. Final report

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-01

    The work scope of the RAK-2 project has involved research on quantification of the effects of selected severe accident phenomena for Nordic nuclear power plants, development and testing of a computerised accident management support system and data collection and description of various mobile reactors and of different reactor types existing in the UK. The investigations of severe accident phenomena focused mainly on in-vessel melt progression, covering a numerical assessment of coolability of a degraded BWR core, the possibility and consequences of a BWR reactor to become critical during reflooding and the core melt behavior in the reactor vessel lower plenum. Simulant experiments were carried out to investigate lower head hole ablation induced by debris discharge. In addition to the in-vessel phenomena, a limited study on containment response to high pressure melt ejection in a BWR and a comparative study on fission product source term behaviour in a Swedish PWR were performed. An existing computerised accident management support system (CAMS) was further developed in the area of tracking and predictive simulation, signal validation, state identification and user interface. The first version of a probabilistic safety analysis module was developed and implemented in the system. CAMS was tested in practice with Barsebaeck data in a safety exercise with the Swedish nuclear authority. The descriptions of the key features of British reactor types, AGR, Magnox, FBR and PWR were published as data reports. Separate reports were issued also on accidents in nuclear ships and on description of key features of satellite reactors. The collected data were implemented in a common Nordic database. (au) 39 refs.

  19. Computerised severe accident management aids

    International Nuclear Information System (INIS)

    The OECD Halden Reactor Project in Norway is running two development projects in the area of computerised accident management in cooperation with the Swedish nuclear plant Forsmark unit 2. Also other nuclear organisations in the Nordic countries take part in the projects. The SAS II system is installed at Forsmark and is now being validated against the plant compact simulator and is later to be installed in the plant control room. It is designed to follow all defined critical safety functions in the same manner as is done in the functionally oriented Emergency Operating Procedures. The shift supervisor thus uses SAS II as a complementary information system after a plant disturbance . The plant operators still use the ordinary instrumentation and the event oriented procedures. This gives to a high extent both redundancy and diversity in information channels and in procedures. Further, a new system is under discussion which goes a step further in accident management than SAS II. It is called the Computerised Accident Management Support (CAMS) system. The objective is to make a computerised tool that can assist both the control room crew and the technical support centre in accident mitigation, especially in the early stages of an accident where the integrity of the core still can be maintained if proper counteractions to the accident sequence are taken. In CAMS another approach is taken than in SAS II by putting the process parameters in focus. A more elaborate signal validation is proposed. The validated signals are input to models that calculates mass and energy balances of the primary system. Among parameters calculated are residual heat. Experiences from these two approaches to computerised accident management support are presented and discussed. In summary: The original project proposal aimed particularly for operator and TSC support during severe accidents. In the CAMS design proposal we have, however, promoted the SMABRE code which is not designed for such

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

    International Nuclear Information System (INIS)

    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)

  1. Important severe accident research issues after Fukushima accident

    International Nuclear Information System (INIS)

    After the Fukushima accident several investigation committees issued reports with lessons learned from the accident in Japan. Among those lessons, several recommendations have been made on severe accident research. Similar to the EURSAFE efforts under EU Program, review of specific severe accident research items was started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation has been started since the Fukushima accident. Additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality have been covered. The review results are categorized in ten major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, direct containment heating, recriticality and instrumentation in severe accident conditions. Based on these activities and also author's personal view, the present paper describes the perspective of important severe accident research issues after Fukushima accident. Those are specifically investigation of damaged core and components, advanced severe accident analysis capabilities and associated experimental investigations, development of reliable passive cooling system for core/containment, analysis of hydrogen behavior and investigation of hydrogen measures, enhancement of removal function of radioactive materials of containment venting, advanced instrumentation for the diagnosis of severe accident and assessment of advanced containment design which excludes long-term evacuation in any severe accident situations. (author)

  2. Severe accident management. Prevention and Mitigation

    International Nuclear Information System (INIS)

    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

  3. TRACE/FRAPTRAN analysis of Kuosheng (BWR/6) Nuclear Power Plant for the similar Fukushima accident

    International Nuclear Information System (INIS)

    Kuosheng nuclear power plant (NPP) is a BWR/6 type NPP and located on the northern coast of Taiwan. In order to assess Kuosheng NPP’s safety under the similar Fukushima accident conditions, Kuosheng NPP safety analysis TRACE/SNAP/FRAPTRAN models were established in this research. There were three main steps in this study. The first step was the establishment of Kuosheng NPP TRACE/SNAP model. The next step was the transient analysis of Kuosheng NPP TRACE/SNAP model under the similar Fukushima accident or more severe conditions. Besides, in order to confirm the mechanical property and integrity of fuel rods, the third step was FRAPTRAN analysis. Finally, the animation model of Kuosheng NPP was presented by using animation function of SNAP. TRACE and FRAPTRAN results depicted that zirconium-water reaction generated in SBO or SBO + LOCA transient (no water injected case). It may cause a safety issue in fuel rods. However, if the fire water (39 kg/sec) injected to the vessel at 800 sec in SBO + LOCA transient, TRACE and FRAPTRAN results implied that the integrity of fuel rod was kept. (author)

  4. A Survey of Implementation of Severe Accident Management in Sweden

    International Nuclear Information System (INIS)

    improvements for the BWR's and PWR's have been performed in parallel and following the same principles. However, the transition from BERG to SAMG for the Ringhals PWR's means a great improvement. Therefore, future work on development of strategies and documentation for severe accident management will probably focus on the BWR's. (author)

  5. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

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

  6. Effect of thermal-hydraulic feedback on the BWR rod drop accident

    International Nuclear Information System (INIS)

    An important design-basis accident for boiling water reactors (BWR's) is the rod drop accident (RDA). This accident is defined to be a rapid reactor transient caused by an accidental drop (out of the core) of the highest-worth control rod at various conditions ranging from cold start-up to about 10% of rated power. For most BWR designs the highest worth rod is normally situated at the center of the core. Despite the fact that the chance of a RDA in extremely unlikely, the consequence of the RDA is of concern because of the potential for damage to fuel rods. Neglecting moderator feedback during the RDA is a poor assumption because energy is deposited in the fuel over a 3 to 4 second time period and hence there is time for heat to be conducted to the coolant. This may tend to ameliorate the accident considerably. Evaluation of the thermal-hydraulic feedback effect on the RDS in a BWR has been scarce in the literature. The object of this paper is to demonstrate the beneficial effect of thermal-hydraulic feedback in the RDA

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    CrAl would tend to generate heat and hydrogen from oxidation at a slower rate compared to the zirconium-based alloys in use today. The previous study, [2], of the FeCrAl ATF concept during station blackout (SBO) severe accident scenarios in BWRs was based on simulating short term SBO (STSBO), long term SBO (LTSBO), and modified SBO scenarios occurring in a BWR-4 reactor with MARK-I containment. The analysis indicated that FeCrAl had the potential to delay the onset of fuel failure by a few hours depending on the scenario, and it could delay lower head failure by several hours. The analysis demonstrated reduced in-vessel hydrogen production. However, the work was preliminary and was based on limited knowledge of material properties for FeCrAl. Limitations of the MELCOR code were identified for direct use in modeling ATF concepts. This effort used an older version of MELCOR (1.8.5). Since these analyses, the BWR model has been updated for use in MELCOR 1.8.6 [10], and more representative material properties for FeCrAl have been modeled. Sections 2 4 present updated analyses for the FeCrAl ATF concept response during severe accidents in a BWR. The purpose of the study is to estimate the potential gains afforded by the FeCrAl ATF concept during BWR SBO scenarios.

  8. MAAP BWR application guidelines

    International Nuclear Information System (INIS)

    The MAAP Thermal-Hydraulic Qualification and Application Project has as its objective to identify those thermal-hydraulic phenomena modeled in MAAP which are important in predicting severe accident sequences, to qualify those models and to provide guidelines for use of the code. This report provides user guidelines for use of the BWR version of MAAP. The report includes a discussion of the important features of the BWR that are modeled in MAAP, the MAAP modeling of phenomena important to predicting severe accidents and user guidelines for several accident sequences

  9. CANDU safety under severe accidents

    International Nuclear Information System (INIS)

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

  10. A framework for assessing severe accident management strategies

    International Nuclear Information System (INIS)

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

  11. Transient boiling and void formation during postulated reactivity-initiated accident in BWR: Experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (author)

  12. Transient boiling and void formation during postulated reactivity initiated accident in BWR: experimental simulation

    International Nuclear Information System (INIS)

    The current safety analysis of the postulated reactivity initiated accident (RIA) in the boiling water reactor (BWR) neglects the favorable effect of voids because of the difficulties in predicting void formation in transient boiling. This paper presents experimental results on the transient void formation in response to a step heating of a surface facing to low-pressure subcooled water. The void fractions are measured by measuring optically the water surface movement or water velocity induced by the void formation. (authors)

  13. Severe accident analysis methodology in support of accident management

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  15. Behavior of small-sized BWR fuel under reactivity initiated accident conditions

    International Nuclear Information System (INIS)

    The present work was performed on this small-sized BWR fuel, where Zr liner and rod prepressurization were taken as experimental parameters. Experiment was done under simulated reactivity initiated accident (RIA) conditions at Nuclear Safety Research Reactor (NSRR) belonged to Japan Atomic Energy Research Institute (JAERI). Major remarks obtained are as follows: (1) Three different types of the fuel rods consisted of (a) Zr lined/pressurized (0.65MPa), (b) Zr lined/non-pressurized and (c) non-Zr lined/pressurized (o.65MPa) were used, respectively. Failure thresholds of these were not less than that (260 cal/g·fuel) described in Japanese RIA Licensing Guideline. Small-sized BWR and conventional 8 x 8 BWR fuels were considered to be in almost the same level in failure threshold. Failure modes of the three were (a) cladding melt/brittle, (b) cladding melt/brittle and (c) rupture by large ballooning, respectively. (2) The magnitude of pressure pulse at fuel fragmentation was also studied by lined/pressurized and non-lined/pressurized fuels. Above the energy deposition of 370 cal/g·fuel, mechanical energy (or pressure) was found to be released from these fragmented fuels. No measurable difference was, however, observed between the tested fuels and NSRR standard (and conventional 8 x 8 BWR) fuels. (3) It is worthy of mentioning that Zr liner tended to prevent the cladding from large ballooning. Non-lined/pressurized fuel tended to cause wrinkle deformation at cladding. Hence, cladding external was notched much by the wrinkles. (4) Time to fuel failure measured from the tested BWR fuels (pressurization < 0.6MPA) was longer than that measured from PWR fuels (pressurization < 3.2MPa). The magnitude of the former was of the order of 3 ∼ 6s, while that of the latter was < 1s. (J.P.N.)

  16. Development of severe accident training support system

    International Nuclear Information System (INIS)

    In order for appropriate decision-making during plant operation and management, the professional knowledge, expert's opinion, and previous experiences as well as information for current status are utilized. The operation support systems such as training simulators have been developed to assist these decision-making process, and most of them cover from normal operation to emergency operation because of the very low frequency of severe accident and of uncertaintics included in severe accident phenomena and scenarios. However, the architectures for severe accident management are being established based on severe accident management guidelines in some developed countries. Recentrly, in Korea, as teh severe accident management guideline was developed, the basis for establishing severe accident management architecture is prepared and this leads to the development of tool for systematic education and training for personnel related to severe accident management. The severe accident taining support system thus is developed to assist decision-making during execution of severe accident management guidelines by providing plant status information, prefessional knowledge for phenomena and scenarios, expected behavior for strategy execution, and so on

  17. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    International Nuclear Information System (INIS)

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed

  18. Current and anticipated use of thermal-hydraulic codes for BWR transient and accident analyses in Japan

    Energy Technology Data Exchange (ETDEWEB)

    Arai, Kenji; Ebata, Shigeo [Toshiba Corp., Yokohama (Japan)

    1997-07-01

    This paper summarizes the current and anticipated use of the thermal-hydraulic and neutronic codes for the BWR transient and accident analyses in Japan. The codes may be categorized into the licensing codes and the best estimate codes for the BWR transient and accident analyses. Most of the licensing codes have been originally developed by General Electric. Some codes have been updated based on the technical knowledge obtained in the thermal hydraulic study in Japan, and according to the BWR design changes. The best estimates codes have been used to support the licensing calculations and to obtain the phenomenological understanding of the thermal hydraulic phenomena during a BWR transient or accident. The best estimate codes can be also applied to a design study for a next generation BWR to which the current licensing model may not be directly applied. In order to rationalize the margin included in the current BWR design and develop a next generation reactor with appropriate design margin, it will be required to improve the accuracy of the thermal-hydraulic and neutronic model. In addition, regarding the current best estimate codes, the improvement in the user interface and the numerics will be needed.

  19. The severe accident research program at KIT

    International Nuclear Information System (INIS)

    The understanding of the plant behaviour under beyond design basis accidents as well as the interaction of the operators with the plant is the most important prerequisite to develop proper strategies to both control the accident progression and to minimize the radiological risk that may derive from operating nuclear power plants. In view of the Fukushima accident, a review of many issues important to safety e.g. severe accident analysis methodologies and assumptions, emergency operational procedures, severe accident management procedures (SAM), decision lines of the emergency team, etc. is needed to draw conclusions in order to avoid a repetition of Fukushima-like accidents.In addition, situations like the ‘black control room’ need to be reconsidered and a re-evaluation of the necessary instrumentation for hypothetical severe accident situations is urgently needed. If the real plant state during core meltdown accidents is unknown, no effective measures can be initiated by the emergency team in order to assure the integrity of the safety barriers and hence the release of radioactive material to the environment. The work performed in this area is integrated in the European Networks such as SARNET (Severe Accident Research Network) for the severe accidents, and for emergency management in the NERIS-TP. In future all the activities will be included in the NUGENIA platform. A brief overview of the KIT activities together with the experimental test facilities is given

  20. Severe Accident Research Program plan update

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Alejandro Nuñez-Carrera; Raúl Camargo-Camargo; Gilberto Espinosa-Paredes; Adrián López-García

    2012-01-01

    The objective of this paper is the simulation and analysis of the BoilingWater Reactor (BWR) lower head during a severe accident. The COUPLE computer code was used in this work to model the heatup of the reactor core material that slumps in the lower head of the reactor pressure vessel. The prediction of the lower head failure is an important issue in the severe accidents field, due to the accident progression and the radiological consequences that are completely different with or without the...

  2. Monitoring severe accidents using AI techniques

    International Nuclear Information System (INIS)

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

  3. KTH experiments on severe accident phenomena relevant to Swedish BWRs

    International Nuclear Information System (INIS)

    A significant fraction of national electricity production in Sweden is supplied by nuclear power plants with BWR reactors. Severe accident management concept of Swedish BWRs, which has been developed in 70-80s, envisages ex-vessel fragmentation and quenching of corium melt in a deep water-filled reactor pit and long term coolability of corium debris bed with water natural convection through the bed open porosity. The paper deals with experimental studies of several severe accident phenomena critical for the chosen SAM strategy: Corium melt jet fragmentation in water; debris bed formation and its properties; debris bed evolution and particle spreading; debris bed coolability; steam explosion during FCI in stratified configuration. Some observations, results and main conclusions from the listed experiments with high temperature corium simulants are presented. The experimental data were used for development and validation of different models and tools, such as MEWA and DECOSIM simulating melt arrest and coolability at the late phase of severe accident under quench and boil-off conditions. The studies were carried out in the Division of Nuclear Power Safety at the Royal Institute of Technology (KTH), Stockholm in the frames of different national and international projects and programs supported by industry, regulators, research organizations and EU. The EU part of the research was coordinated by the SARNET network. The SARNET collaboration will be continued in NUGENIA format. (author)

  4. Iodine behaviour in severe accidents

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-12-01

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

  5. Severe accidents in Nuclear Power Plants

    International Nuclear Information System (INIS)

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

  6. SAMSON: Severe Accident Management System Online Network

    International Nuclear Information System (INIS)

    SAMSON, Severe Accident Management System Online Network, is a computational tool used in the event of a nuclear power plant accident by accident managers in the Technical Support Centers (TSC) and Emergency Offsite Facilities (EOF). 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. The status points analyzed include radiation levels, flow rates, pressure levels, temperatures and water levels. SAMSON uses an expert system as well as neural networks trained with the back propagation learning algorithm to make predictions. Previous training on data from accident analysis code 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 (SGTR), with breaks ranging from one tube to eights tubes, and loss of coolant accidents (LOCA), with breaks ranging from 0.001 square feet in size to breaks 3.0 square feet. SAMSON contains several neural networks for each accident type and break size, and chooses the correct network after accident classification by in expert system. SAMSON also provides information concerning the status of plant sensors and recovery strategies

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

    International Nuclear Information System (INIS)

    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)

  8. Thermal Response of the 44-BWR Waste Package to a Hypothetical Fire Accident

    International Nuclear Information System (INIS)

    The purpose of this calculation is to determine the thermal response of the 44-boiling water reactor (BWR) waste package (WP) to the hypothetical regulatory fire accident. The objective is to calculate the temperature response of the waste package materials to the hypothetical short-term fire defined in 10 CFR 7 1, Section 73(c)(4), Reference 1. The scope of the calculation includes evaluation of the accident with the waste package above ground, at the Yucca Mountain surface facility. The scope of this calculation is limited to the two-dimensional waste package temperature calculations to support the waste package design. The information provided by the sketches attached to this calculation is that for the potential design of the type of WP considered in this calculation. In addition to the nominal design configuration thermal load case, the effects of varying the BWR thermal load are determined. The associated activity is the development of engineering evaluations to support the Licensing Application (LA) design activities

  9. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Muhammad Hashim

    2013-04-01

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

  11. Deterministic analyses of severe accident issues

    International Nuclear Information System (INIS)

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

  12. Containment leakage during severe accident conditions

    International Nuclear Information System (INIS)

    An alternate to the THRESHOLD model used in most severe accident risk assessments has been investigated. One reference plant for each of six containment types has been studied to determine the magnitude of containment leakage that would result from the pressures and temperatures associated with severe accident conditions. Containment penetrations having the greatest potential for early containment leakage are identified. The studies indicate that containment leakage through penetrations prior to reaching containment threshold pressures (currently reported containment shell failure pressures) should be considered in severe accident risk assessments. Failure of non-metallic seals for containment penetrations can be a significant source of containment leakage under severe accident pressure and temperature conditions. Although studies of containment types are useful in identifying sources of containment leakage, final conclusions may need to be plant specific. Recommendations concerning future studies to better develop the use of continuous leakage models are provided. 9 references, 4 figures, 2 tables

  13. Severe accident management: a summary of the VAHTI and ROIMA projects

    International Nuclear Information System (INIS)

    Two severe accident research projects: 'Severe Accident Management' (VAHTI), 1994-96 and 'Reactor Accidents' Phenomena and Simulation (ROIMA) 1997-98. have been conducted at VTT Energy within the RETU research programme. The main objective was to assist the severe accident management programmes of the Finnish nuclear power plants. The projects had several subtopics. These included thermal hydraulic validation of the APROS code, studies of failure mode of the BWR pressure vessel, investigation of core melt progression within a BWR pressure vessel, containment phenomena, development of a computerised severe accident training tool, and aerosol behaviour experiments. The last topic is summarised by another paper in the seminar. The projects have met the objectives set at the project commencement. Calculation tools have been developed and validated suitable for analyses of questions specific for the Finnish plants. Experimental fission product data have been produced that can be used to validate containment aerosol codes. The tools and results have been utilised in plant assessments. One of the main achievements has been the computer code PASULA for analysis of interactions between core melt and pressure vessel. The code has been applied to pressure vessel penetration analysis. The results have shown the importance of the nozzle construction. Modelling possibilities have recently improved by addition of a creep and porous debris models. Cooling of a degraded BWR core has been systematically studied as joint Nordic projects with a set of severe accident codes. Estimates for coolable conditions have been provided. Recriticality due to reflooding of a damaged core has been evaluated. (orig.)

  14. Conclusions on severe accident research priorities

    International Nuclear Information System (INIS)

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

  15. Severe accident analysis using dynamic accident progression event trees

    Science.gov (United States)

    Hakobyan, Aram P.

    In present, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. One of the principal deficiencies lies in the static nature of conventional APETs. In the conventional event tree techniques, the sequence of events is pre-determined in a fixed order based on the expert judgments. The main objective of this PhD dissertation was to develop a software tool (ADAPT) for automated APET generation using the concept of dynamic event trees. As implied by the name, in dynamic event trees the order and timing of events are determined by the progression of the accident. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. The function of a dynamic APET developed includes prediction of the conditions, timing, and location of containment failure or bypass leading to the release of radioactive material, and calculation of probabilities of those failures. Thus, scenarios that can potentially lead to early containment failure or bypass, such as through accident induced failure of steam generator tubes, are of particular interest. Also, the work is focused on treatment of uncertainties in severe accident phenomena such as creep rupture of major RCS components, hydrogen burn, containment failure, timing of power recovery, etc. Although the ADAPT methodology (Analysis of Dynamic Accident Progression Trees) could be applied to any severe accident analysis code, in this dissertation the approach is demonstrated by applying it to the MELCOR code [1]. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a

  16. A review of severe accident assessment

    International Nuclear Information System (INIS)

    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.)

  17. 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).

  18. The development of severe accident analysis technology

    International Nuclear Information System (INIS)

    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)

  19. 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

    in assessment of plant safety. This work package was also designed to distinguish the differences between the attitudes and approaches followed by the various regulatory organisations in Europe, Eastern Europe, USA and Japan. Work Package 5: Relevance of example PSA results to SA research. The objective of their work package was to employ the results of some recent PSAs (preferably for a PWR and a BWR) and relate their findings to the results obtained in SA research, and to the effectiveness of the SAM measures already taken or contemplated. Work Package 6: The state of resolution of the SA issues with respect to the needs. The objective of this work package is to have another look at the state of the resolution of the severe accident issues which have been identified over the years, and relate that to what the needs of the regulatory organizations are in terms of their functions. Work Package 7: Regulatory use of the results of severe accident research. The objective is to identify the results of the SA research which the regulatory organizations, over the years, have used in either defining specific regulatory actions or in not taking specific actions. Work Package 8: Remaining issues and concerns. The objective of the work here is to review the work in the previous work package and identify what are the remaining unresolved safety issues and concerns for which sufficient results of the SA research are not available. Work Package 9: Recommendations on future directions of severe accident research. The purpose of this work package is to provide recommendations to E.U. (and to the readers) by the authors of this report on the directions that should be followed, in the future for the conduct of severe accident research. These recommendations are in essence the conclusions of this study.

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

    International Nuclear Information System (INIS)

    in assessment of plant safety. This work package was also designed to distinguish the differences between the attitudes and approaches followed by the various regulatory organisations in Europe, Eastern Europe, USA and Japan. Work Package 5: Relevance of example PSA results to SA research. The objective of their work package was to employ the results of some recent PSAs (preferably for a PWR and a BWR) and relate their findings to the results obtained in SA research, and to the effectiveness of the SAM measures already taken or contemplated. Work Package 6: The state of resolution of the SA issues with respect to the needs. The objective of this work package is to have another look at the state of the resolution of the severe accident issues which have been identified over the years, and relate that to what the needs of the regulatory organizations are in terms of their functions. Work Package 7: Regulatory use of the results of severe accident research. The objective is to identify the results of the SA research which the regulatory organizations, over the years, have used in either defining specific regulatory actions or in not taking specific actions. Work Package 8: Remaining issues and concerns. The objective of the work here is to review the work in the previous work package and identify what are the remaining unresolved safety issues and concerns for which sufficient results of the SA research are not available. Work Package 9: Recommendations on future directions of severe accident research. The purpose of this work package is to provide recommendations to E.U. (and to the readers) by the authors of this report on the directions that should be followed, in the future for the conduct of severe accident research. These recommendations are in essence the conclusions of this study

  1. Severe accident management. Optimized guidelines and strategies

    International Nuclear Information System (INIS)

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

  2. Containment severe accident management - selected strategies

    International Nuclear Information System (INIS)

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

  3. Severe accident risks from external events

    Institute of Scientific and Technical Information of China (English)

    Randall O Gauntt

    2013-01-01

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

  4. PHEBUS reactor: The driving of a severe accident

    International Nuclear Information System (INIS)

    The experimental PHEBUS FP (Fission Products) programme is dedicated to the study of core meltdown in a light water reactor (PWR or BWR) during a severe accident and to the behaviour of released fission products. For that purpose, a core meltdown accident is reproduced on a reduced scale, the core being simulated by a bundle of fuel rods or a debris bed). The objectives of each Fission Product Test are expressed in term of fuel degradation, FPs release and transport characteristics. Those objectives fix the specific fuel characteristic (irradiated or not, bundle or debris bed geometry) and thermal-hydraulic conditions for each tests. The test elaboration is based on main steps (definition of all the probable scenarii for fuel degradation, pre-calculations, definition of the test device instrumentation,...) which led to the test conduct strategy with respect to the lessons learnt from the previous tests. This test conduct strategy is based on several shutdown conditions which enables to reach the experimental objective, inside the safety limits. (author)

  5. Severe accident issue resolution -- definition and perspective

    International Nuclear Information System (INIS)

    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

  6. Severe accident management concept for LWRS

    International Nuclear Information System (INIS)

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

  7. Application of FFTBM to severe accidents

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

  9. Source term formation in CANDU severe accidents

    International Nuclear Information System (INIS)

    The paper presents the phenomena involved in the most important CANDU severe accident (LOCA+LOECC, SBO, SGTR, EFF). Fission products are grouped in classes taking into consideration the half time, volatility, chemistry and biological activity. An analysis of the paths on which the release of the fission products to the environment occurs is performed. For each type of CANDU severe accident the process of source term formation, the magnitude and structure of source term and also the timing are presented on the basis of SOPHAEROS, CPA and IODE (modules included in ASTEC code) calculations, completed with literature results. The discussion about the involved sources of uncertainties is also presented taking into account the complexity of phenomena, the great number of parameters and limited availability of experimental data. Some general recommendations are developed in order to use the results in achieving the procedures for protective actions during a reactor accident. (authors)

  10. Severe Accident Test Station Activity Report

    Energy Technology Data Exchange (ETDEWEB)

    Pint, Bruce A [ORNL; Terrani, Kurt A [ORNL

    2015-06-01

    Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

  11. Application of the WECHSL code to PWR and BWR specific accident scenarios

    International Nuclear Information System (INIS)

    The WECHSL Mod3 version is used to perform an accident analysis for a 1300 MW PWR and a BWR. The analysis starts after the melt has penetrated the reactor pressure vessel and is contained in the dry reactor cavity. The initial melt temperature is estimated to be 2673 K. In the initial phase of the melt/concrete interaction, the dominant energy source in the melt is the energy released in the zirconium oxidation reactions with the concrete decomposition products. Hence the concrete composition will determine the Zr-oxidation and the gas release rates as well as the composition of the released gases. Recent experiments and analyses have shown that the solidus temperature of the oxidic melt decreases much more rapidly with addition of concrete oxide than modelled previously. The solidus temperature of the oxide phase drops rapidly as concrete oxides are incorporated into the melt, approaching the concrete solidus at only about 10 to 20 weight percent of concrete oxides. The calculations are performed using the old estimate and the new solidus temperatures for both reactor types in order to study the influence of that oxide solidus temperature. The condensed Zr/SiO2 chemistry is only relevant for the PWR because of the high content of SiO2 in the siliceous concrete basemat. Compared to former analyses for the PWR the much faster zirconium oxidation leads to a higher temperature of about 100 K in the early phase of melt/concrete interaction and therefore the crust formation process starts later than in the former analyses leading to a longer duration of high gas release rates dominated by H2 because of more effective heat transfer to the concrete in this period of time. The concrete basemat of the BWR consists of pure limestone with a decomposition temperature which is higher than the solidus temperature of the metallic melt. This high concrete decomposition temperature prevents a crust formation at the metal-concrete boundary. Hence a very efficient heat transfer leads

  12. Fission product behaviour in severe accidents

    International Nuclear Information System (INIS)

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

  13. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Occupational Radiation Protection in Severe Accident Management

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Severe accident aerosol research in Finland

    International Nuclear Information System (INIS)

    The retention of fission products in the steam generator tubing and in the secondary side is poorly understood at the moment. Most experimental programs have concentrated on the initial stages of deposition. Much less attention has been paid to the situations when deposition-resuspension-revaporisation are important as the deposit layers are getting thicker. The understanding of fission product deposition in realistic steam generator conditions is needed to design efficient accident management procedures. For example if there is large deposition already in the ruptured pipe(s), the accident management procedure is different from the case where most deposition would occur in the secondary side. This is considered very important because steam generator tube rupture sequences are included in the risk dominant sequences. Aerosol deposition has been studied widely in laboratory scale. However, most of the studies have concentrated on situations where the deposit layer is thin and do not significantly affect the process. In severe accident applications the most important deposition studies have been LACE, STORM, TUBA, TRANSAT and AIDA programmes. None of these tests considered steam generator conditions. Thus we can say that there is basic knowledge on aerosol deposition and removal from gas streams in water pools, but it can not be applied directly to steam generator tube rupture cases. At the moment the effectiveness of such accident management procedures as secondary side flooding can not be verified as there is no experimental data and the models in severe accident codes are poor or non-existing. As a results of this work we will get data on deposition in the tubing, in the break location and in the secondary side. Experiments will be performed in horizontal steam generators (VVER reactors). (orig.)

  20. Severe accident research in the UK

    International Nuclear Information System (INIS)

    Severe Accident R and D in the UK builds on more than 25 years experience and for the PWR is firmly committed to international collaboration. The focus for the work has been the support for a comprehensive Level 3 PSA for Sizewell 'B'. The paper outlines the particular contributions that the UK has made to research in direct containment heating, steam explosions, fission product behaviour and code development and assessment. (author)

  1. Nuclear power plant Severe Accident Research Plan

    International Nuclear Information System (INIS)

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

  2. Severe Accident Test Station Design Document

    Energy Technology Data Exchange (ETDEWEB)

    Snead, Mary A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Yan, Yong [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Howell, Michael [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Keiser, James R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrani, Kurt A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-01

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

  3. Current severe accident research facilities and projects

    International Nuclear Information System (INIS)

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

  4. Development of a totally integrated severe accident training system

    International Nuclear Information System (INIS)

    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)

  5. Sarnet lecture notes on nuclear reactor severe accident phenomenology

    International Nuclear Information System (INIS)

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

  6. 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.

  7. Severe radiation accidents and the environment

    International Nuclear Information System (INIS)

    In severe radiation accidents with releases of radionuclides into the environment, high enough radiation doses are formed to potentially cause direct radiation injury of natural ecosystems. The dose fields characterizing the irradiation of plants, animals and humans in radioactive contamination of the environment are highly heterogeneous, and irradiation of natural objects per se has features such as non-equidosal effects. In other words, humans and various objects in the natural environment can receive different absorbed doses for an equal density of radioactive fallout. 5 refs

  8. Severe Accidents in the Energy Sector

    International Nuclear Information System (INIS)

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

  9. Severe Accidents in the Energy Sector

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-11-01

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

  10. Experimental investigations of BWR pressure suppression pool behavior under loss of coolant accident conditions

    International Nuclear Information System (INIS)

    The experiments discussed in this paper look into different processes which may occur during a loss-of-coolant accident in the pressure suppression pool of a Boiling Water Reactor (BWR). These processes include: a) development of a thermal stratification, b) bubble dynamics and related water flow during continuous release of air and c) air blowdown and associated water slug phenomenon in the water pool. The experiments have been performed in the THAI test facility, which is a cylindrical vessel of 9.2 m height, 3.2 m diameter and with a gas volume of 60 m3. The variation in the investigated test parameters included, steam and air mass flux, initial water pool temperature, blowdown pressures, downcomer submergence, etc. A systematic variation of the test parameters allowed better understanding of the phenomena. Experiments discussed in this paper were performed with a vertical downcomer of 0.1 m diameter and 2 m submergence depth in the water pool. For the blowdown experiments, a separate interconnecting vessel of 1 m3 volume was used to inject air at pressures between 3 bar and 10 bar. A high speed camera (1000 fps) was installed to visualize the formation and propagation of air bubbles in the suppression pool and the resulting pool swelling phenomena. Customized instrumentation applied during the tests included grids of densely spaced thermocouples and of pressure transducers at various locations in order to capture the temperature distribution in the pool and the water slug induced pressure loadings, respectively. The present paper discusses the main outcome of the selected experiments. On the whole the experimental data may be very useful for code validation. (authors)

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

    DEFF Research Database (Denmark)

    Prato, Carlo Giacomo; Kaplan, Sigal

    2014-01-01

    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...

  12. Several accidents about ERHRS of CEFR

    International Nuclear Information System (INIS)

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

  13. New regulatory standards and severe accident countermeasures at nuclear power plants

    International Nuclear Information System (INIS)

    Following the accident of Tokyo Electric Power Company's Fukushima Daiichi Nuclear Power Station, the basic policy for the cause for the escalation of severe accidents was shown, and based on this, new regulatory standards were enacted in July 2013. In addition to the above countermeasures, these standards intensified countermeasures for natural disasters such as tornado and volcanic eruption, as well as for fire and flooding inside power plants, and aircraft terrorism. The standards demanded that even if a fire has occurred, the cooling and monitoring of a nuclear power plant should be continued, using a designated severe accident countermeasure facility. For PWR, the installation of filter vents and designated severe accident countermeasure facility are given a grace period of five years. For BWR, measures for the breakage of containment, such as filter vents, and emission prevention measures for radioactive substances are prerequisites for the resumption of the operation. As specific examples of tsunami countermeasures of power plants across Japan, the following are introduced: (1) seawalls and tide gates at Chubu Electric Power Hamaoka Power Station, (2) seawalls at Chubu Electric Power Shimane Power Station, and (3) watertight structure of the reactor building at Tokyo Electric Power Kashiwazaki Power Station. As the severe accident countermeasures of each nuclear power plant in Japan, points are explained on the basis of the core damage prevention measures at the Units 1 and 2 of Kyushu Electric Power Sendai Power Station that are described in the license application for change of reactor installation. (A.O.)

  14. Code strategy for simulating Severe Accident Scenario

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Perspective on post-Fukushima severe accident research

    International Nuclear Information System (INIS)

    After the Fukushima Daiichi accident in March 2011 several investigation committees issued reports with lessons learned from the accident, in which some recommendations on severe accident research are included. The review of specific severe accident research items had already started before Fukushima accident in working group of Atomic Energy Society of Japan (AESJ) in terms of significance of consequences, uncertainties of phenomena and maturity of assessment methodology. Re-investigation started after the Fukushima accident in this working group to cover additional effects of Fukushima accident, such as core degradation behaviors, sea water injection, containment failure/leakage and re-criticality. The review results are categorized in nine major fields; core degradation behavior, core melt coolability/retention in containment vessel, function of containment vessel, source term, hydrogen behavior, fuel-coolant interaction, molten core concrete interaction, recriticality and instrumentation in severe accident conditions. In January 2012, in collaboration with this working group, Research Expert Committee on Evaluation of Severe Accident was established in AESJ in order to investigate severe accident related issues for future LWR development. Based on these activities and also author's personal view, the present paper describes the seven important severe accident research issues after Fukushima accident. They are (1) investigation of damaged core and components, (2) advanced severe accident analysis capabilities and associated experimental investigations, (3) development of reliable passive cooling system for core/containment, (4) analysis of hydrogen behavior and investigation of hydrogen measures, (5) enhancement of removal function of radioactive materials of containment venting, (6) advanced instrumentation for the diagnosis of severe accident and (7) assessment of advanced containment design which exchides long-term evacuation in any severe accident situations

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

    International Nuclear Information System (INIS)

    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)

  18. Aerosol transport in severe reactor accidents

    International Nuclear Information System (INIS)

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

  19. SARNET: Severe accident research network of excellence

    International Nuclear Information System (INIS)

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

  20. 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.

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

    International Nuclear Information System (INIS)

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

  2. Fuel gases generation in the primary contention during a coolant loss accident in a nuclear power plant with reactor type BWR

    International Nuclear Information System (INIS)

    During an accident design base of coolant loos, the hydrogen gas can accumulate inside the primary contention as a result of several generation mechanisms among those that are: 1) the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant, 2) the metals corrosion for the solutions used in the emergency cooling and dew of the contention, and 3) the radio-decomposition of the cooling solutions of post-accident emergency. In this work the contribution of each generation mechanism to the hydrogen total in the primary contention is analyzed, considering typical inventories of zirconium, zinc, aluminum and fission products in balance cycle of a reactor type BWR. In the analysis the distribution model of fission products and hydrogen production proposed in the regulator guide 1.7, Rev. 2 of the US NRC was used. The results indicate that the mechanism that more contributes to the hydrogen generation at the end of a period of 24 hours of initiate the accident is the radio-decomposition of the cooling solutions of post-accident emergency continued by the reaction metal-water involving the zirconium of the fuel cladding with the reactor coolant, and lastly the aluminum and zinc oxidation present in the primary contention. However, the reaction metal-water involving the zirconium of the fuel cladding and the reactor coolant is the mechanism that more contributes to the hydrogen generation in the first moments after the accident. This study constitutes the first part of the general analysis of the generation, transport and control of fuel gases in the primary contention during a coolant loss accident in BWRs. (Author)

  3. Current position on severe accident phenomena

    International Nuclear Information System (INIS)

    The phenomena addressed in this lecture are: in-vessel and ex-vessel hydrogen generation; in-vessel and in-containment natural circulation, steam explosions, direct containment heating, core-concrete interaction; debris coolability, containment strength/failure. The following events were modeled: axial and radial power distribution, two-phase level in the core, steam generation in covered section, decay heat generation, convection to gas, cladding oxidation, cold ballooning and rupture, natural circulation between the core and upper plenum, hydrogen generation, core meltdown, reflooding. Differences between PWR and BWR type reactors

  4. ACR-1000: Enhanced response to severe accidents

    International Nuclear Information System (INIS)

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

  5. Simulation of severe accident in reactor core for training and accident management

    International Nuclear Information System (INIS)

    An Advanced Real-time Severe Accident Simulation (ARTSAS) train reactor operators and accident management teams for scenarios simulating severe accidents in nuclear reactors. The code has been integrated with the real-time tools and the RAINBO graphic package to provide training and analysis tools on workstations as well as on full-scope simulators. (orig.) (4 refs., 1 fig.)

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  9. Iodine chemical forms in LWR severe accidents

    International Nuclear Information System (INIS)

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

  10. On severe accident hydrogen behaviour in Loviisa

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  12. Analysis and research status of severe core damage accidents

    International Nuclear Information System (INIS)

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

  13. 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)

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

    International Nuclear Information System (INIS)

    This report deals with the comparative assessment of accidents risks characteristic for the various electricity supply options. A reasonably complete picture of the wide spectrum of health, environmental and economic effects associated with various energy systems can only be obtained by considering damages due to normal operation as well as due to accidents. The focus of the present work is on severe accidents, as these are considered controversial. By severe accidents we understand potential or actual accidents that represent a significant risk to people, property and the environment and may lead to large consequences. (author)

  15. Possible Methods to Estimate Core Location in a Beyond-Design-Basis Accident at a GE BWR with a Mark I Containment Stucture

    International Nuclear Information System (INIS)

    It is difficult to track to the location of a melted core in a GE BWR with Mark I containment during a beyond-design-basis accident. The Cooper Nuclear Station provided a baseline of normal material distributions and shielding configurations for the GE BWR with Mark I containment. Starting with source terms for a design-basis accident, methods and remote observation points were investigated to allow tracking of a melted core during a beyond-design-basis accident. The design of the GE BWR with Mark-I containment highlights an amazing poverty of expectations regarding a common mode failure of all reactor core cooling systems resulting in a beyond-design-basis accident from the simple loss of electric power. This design is shown in Figure 1. The station blackout accident scenario has been consistently identified as the leading contributor to calculated probabilities for core damage. While NRC-approved models and calculations provide guidance for indirect methods to assess core damage during a beyond-design-basis loss-of-coolant accident (LOCA), there appears to be no established method to track the location of the core directly should the LOCA include a degree of fuel melt. We came to the conclusion that - starting with detailed calculations which estimate the release and movement of gaseous and soluble fission products from the fuel - selected dose readings in specific rooms of the reactor building should allow the location of the core to be verified.

  16. Strategy for the Development of Severe Accident Analysis Technology

    International Nuclear Information System (INIS)

    To ensure the safety of people living near the nuclear power plants during the postulated events of severe accidents, a severe accident management strategy is prepared for the operating reactors and dedicated engineered features for the severe accidents are under research and development for the new reactors, such as GEN-III reactors. To accomplish these tasks, not only a proper understanding of fundamental physics of severe accident phenomena but also reliable computer codes for analyzing the severe accident phenomena is very necessary. This report deals with a strategic plan for a development and provision of computer code system for analyzing the severe accidents. This reports includes a summary of major phenomena of severe accidents, an peer review of the computer codes for analyzing the integral behavior of severe accident scenario and computer codes for analyzing the specific phenomena. Finally, a strategic plan for an equipment of severe accident computer codes either by use of already available computer codes or a development of our own computer codes, which could be competitive with world class foreign computer codes

  17. Analysis of severe accidents in pressurized heavy water reactors

    International Nuclear Information System (INIS)

    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

  18. Experimental data report for Test TS-1 Reactivity Initiated Accident Test in NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    This report presents experimental data for Test TS-1 which was the first in a series of tests, simulating Reactivity Initiated Accident (RIA) conditions using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in October, 1989. Test fuel rod used in the Test TS-1 was a short-sized BWR (7 x 7) type rod which was fabricated from a commercial rod provided from Tsuruga Unit 1 power reactor. The fuel had an initial enrichment of 2.79 % and burnup of 21.3 GWd/t (bundle average). Pulse irradiation was performed at a condition of stagnant water cooling, atmospheric pressure and ambient temperature using a newly developed double container-type capsule. Energy deposition of the rod in this test was evaluated to be about 61 cal/g·fuel (55 cal/g·fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, fuel burnup measurements, transient behavior of the test rod during pulse irradiation and results of post pulse irradiation examinations are contained in this report. (author)

  19. Behavior of irradiated BWR fuel under reactivity-initiated-accident conditions. Results of tests FK-1, -2 and -3

    International Nuclear Information System (INIS)

    Boiling water reactor (BWR) fuel rods with burnups of 41 to 45 GWd/tU were pulse-irradiated in the Nuclear Safety Research Reactor (NSRR) to investigate the fuel behavior during a reactivity initiated accident (RIA) at cold startup. BWR fuel segment rods of 8 x 8BJ (STEP I) type from the Fukushima Daiichi Nuclear Power Station Unit 3 were refabricated into short test rods, and they were subjected to prompt enthalpy insertion from 293 to 607 J/g (70 to 145 cal/g) within about 20 ms. The fuel cladding had enough ductility against the prompt deformation due to pellet cladding mechanical interaction. The plastic hoop strain reached 1.5% at the peak location. The cladding surface temperature locally reached about 600 degC. Recovery of irradiation defects in the cladding due to high temperature during the pulse irradiation was indicated via X-ray diffractometry. The amount of fission gas released during the pulse irradiation was from 3.1% to 8.2% of total inventory, depending on the peak fuel enthalpy and the normal operation conditions. (author)

  20. A Methodology for Evaluating Severe Accident Management Strategies

    International Nuclear Information System (INIS)

    Severe accidents are defined as those which entail at least an initial core damage, in many cases specified as the overcoming of the regulatory fuel. After Fukushima accident, the effectiveness of the severe accident management strategy has been attracted worldwide. There is a typical example of severe accident management strategy like Severe Accident Management and Guideline (SAMG). Unfortunately, suitable method for evaluating the accident management strategy is absence until now. In this study, the evaluation methodology which utilizes the decision tree is developed to evaluate the severe accident management strategies. In addition, we applied the developed methodology to ShinKori nuclear power plant Unit 3, 4 and modeled decision tree for evaluation. In this study, we developed a methodology to evaluate the severe accident management strategy by using decision tree. In addition, the evaluation was carried out by selecting the cavity flooding strategy. Shinkori unit 3, 4 which is APR1400 is selected and analyzed for reference plant. In order to evaluation, decision tree for cavity flooding is modeled. With reliability data, quantification will be conducted. The utility of other severe accident management strategies can be evaluated with proposed methodology in this study. Finally, it is expected that this methodology improves the safety of nuclear power plant

  1. Survey of severe accident experiments and analyses in Japan

    International Nuclear Information System (INIS)

    An overview of Japanese activities in the field of Light Water Reactor (LWR) severe accident experiments and analyses is presented, covering various fields and topics of experimental investigation on severe accident phenomena such as fuel damage and melt progression, fission products release and transport, and component and containment integrity. The current status of analytical investigations on severe accidents is also described in the fields of the level-1 and level-2 (PSA) probabilistic safety assessment studies, code development and assessment activities. Basic considerations for accident management are summarized. (author)

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

    OpenAIRE

    Tusheva, Polina

    2013-01-01

    Accident conditions involving significant core degradation are termed severe accidents /IAEA: NS-G-2.15/. Despite the low probability of occurrence of such events, the investigation of severe accident scenarios is an important part of the nuclear safety research. Considering a hypothetical core melt down scenario in a VVER-1000 light water reactor, the early in-vessel phase focusing on the thermal-hydraulic phenomena, and the late in-vessel phase focusing on the melt relocation into the re...

  3. Use of probabilistic safety analyses in severe accident management

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

  5. Use of an influence diagram and fuzzy probability for evaluating accident management in a BWR

    International Nuclear Information System (INIS)

    This paper develops a new approach for evaluating severe accident management strategies. At first, this approach considers accident management as a decision problem (i.e., ''applying a strategy'' vs. ''do nothing'') and uses influence diagrams. This approach introduces the concept of a ''fuzzy probability'' in the evaluation of an influence diagram. When fuzzy logic is applied, fuzzy probabilities in an influence diagram can be easily propagated to obtain results. In addition, the results obtained provide not only information similar to the classical approach using point-estimate values, but also additional information regarding the impact from imprecise input data. The proposed methodology is applied to the evaluation of the drywell flooding strategy for a long-term station blackout sequence in the Peach Bottom nuclear power plant. The results show that the drywell flooding strategy seems to be beneficial for preventing reactor vessel breach. It is also effective for reducing the probability of the containment failure for both liner melt-through and late overpressurization. Even though there exists uncertainty in the results, ''flooding'' is preferred to ''do nothing'' when evaluated in terms of expected consequences, i.e., early and late fatalities

  6. Research of severe accident induced by small LOCA and accident mitigation

    International Nuclear Information System (INIS)

    Fangjiashan nuclear power plant is modeled, by using MAAP4 code. Base on this model, the small LOCA accident is calculated, which will cause the worst consequence. The response of the plant and relevant severe accident phenomena are obtained. The phenomena of DCH (direct containment heat) happened during the accident, containment failure and release of the fission production are analyzed. Then, according to the related severe accident management and characteristic of this accident, the strategy of mitigating the accident consequence is studied and calculated. The result indicated that the mitigation action is very efficient. Therefore, a feasible strategy of mitigating the severe accident consequence is provided for the three-loop plant like Fangjiashan in China. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

    Lin, C.C.; Lehner, J.R.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Drouin, M. [Nuclear Regulatory Commission, N. Bethesda, MD (United States)

    1995-12-31

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

  8. Severe Accident Management Strategy for EU-APR1400

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-02-01

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

  10. An overview of selected severe accident research and applications

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

  15. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. In-vessel core degradation in LWR severe accidents: a state of the art report to CSNI january 1991

    International Nuclear Information System (INIS)

    This state of the art report on in-vessel core degradation has been produced at the request of CSNI Principal Working Group 2. The objective of the report is to present to CSNI member countries the status of research and related information on in-vessel degraded core behaviour in both Pressurised Water Reactors (PWR) and Boiling Water Reactors (BWR). Information on experiments, codes and comparisons of calculations with experiments up to january 1991 is summarised and reviewed. Integrated codes, which are wider in scope than just in-vessel degradation are covered as well as specialist, degraded core codes. Implications for PWR and BWR plant calculations are considered. Conclusions and recommendations for research, plant calculations and further CSNI activity in this area are the subject of the final chapter. A major conclusion of the report is that early phase core degradation is relatively well understood. However, codes need further development to bring them up to date with the experimental database, particularly to include low temperature liquefaction processes. These processes significantly affect early phase core degradation and their neglect could affect assessments of accident management actions (including recriticality in BWR severe accidents)

  18. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    International Nuclear Information System (INIS)

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP ampersand S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP ampersand S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP ampersand S configuration are given

  19. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

    Energy Technology Data Exchange (ETDEWEB)

    Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

    1995-03-01

    To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

  20. The role of nuclear reactor containment in severe accidents

    International Nuclear Information System (INIS)

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

  1. Small scale BWR core debris eutectics formation and melting experiment

    International Nuclear Information System (INIS)

    A small scale experiment has recently been performed at Oak Ridge under the auspices of the BWR Severe Accident Technology (BWRSAT) program to provide information concerning the formation of mixtures during heatup of representative BWR reactor vessel bottom head debris and to determine the composition and melting temperatures of these mixtures. The initial structure of the bottom head debris layers modeled in the experiment was taken from the results of recent BWR Accident Response (BWRSAR) code predictions for the short-term station blackout accident sequence. The experimental results provide useful information concerning the mixtures formed and their proportions and properties. The observed run-off of a stainless steel-zirconium eutectic alloy supports the contention that the initial pour from a BWR reactor vessel would consists of molten metals at relatively low temperatures. (orig.)

  2. Application of PCTRAN-3/U to studying accident management during PWR severe accident

    International Nuclear Information System (INIS)

    In order to improve the safety of nuclear power plant, operator action should be taken into account during a severe accident. While it takes a long time to simulate the plant transient behavior under a severe accident in comparison with the design based accident, a transient simulator should have both high speed calculation capability and interactive functions to model the operating procedures. PCTRAN has been developing to be a simple simulator by using a personal computer to simulate plant behavior under an accident condition. While currently available means usually take relatively long time to simulate plant behavior, using a current high-powered personal computer (PC), PCTRAN-3/U code is designed to operate at a speed significantly faster than real-time. The author describes some results of PCTRAN application in studying the efficiency of accident management for a pressurized water reactor (PWR) during an severe accident

  3. Using MARS to assist in managing a severe accident

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. The philosophy of severe accident management in the US

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. Preliminary severe accident management strategies for Wolsong nuclear power plants

    International Nuclear Information System (INIS)

    Severe accident management strategies for Wolsong 2,3,4 Nuclear Power Plants are presented. The defense in depth concept, which limits release of radioactive materials out of containment building, is applied to develop these strategies. These strategies are actions to prevent or to mitigate core damage, rupture of calandria vessel, rupture of calandria vault, rupture of containment building, and release of radioactive materials. These strategies are deduced from the results of level 2 PSA for Wolsong NPPs. These preliminary results will be assessed further and proved to be effective to Wolsong Plants. Then these severe accident management strategies can be used to develop severe accident management program for Wolsong NPPs

  8. A computational study on instrumentation guide tube failure during a severe accident in boiling water reactors

    International Nuclear Information System (INIS)

    This paper focuses on the nature and timing of Instrumentation Guide Tube (IGT) failure in case of severe core melt accident in a Nordic type Boiling Water Reactor (BWR). First, a 2D structural analysis of a RPV lower head is performed to determine global vessel deformation, timing and mode of failure. Next, a structural analysis is also performed on a 3D IGT section taking into account the influence of global vessel deformation and thermo-mechanical load from the melt pool. We have found that the IG tube was not clamped in the housing at the time when welding ring of the IGT nozzle has been melted and global failure of the vessel wall has not started yet. This suggests that IGT failure is the dominant failure mode in the considered case of a large (~200 tons) melt pool. (author)

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

    International Nuclear Information System (INIS)

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

  10. Severe Accident Management Measures Introduced in Belgian NPP's

    International Nuclear Information System (INIS)

    In response to the Belgian Safety Authorities' request to address the severe accident issue within a decennial safety review, Tractebel, on behalf of the Belgian Utility, Electrabel, examined in detail specific severe accident topics and provided the Utility with several measures that could be implemented to reduce the risk associated with beyond-design accidents. The present paper summarizes the key elements of the approach applied in Belgium: - Presentation of plant-specific studies related to severe accident issues; - Use of PSA results; - Inputs of international R and D projects; - Selection and justification of severe accident measures; - Comparative study between possible mitigative measures; - Definition and justification of implemented severe accident management strategies. The vulnerability to severe accidents as well as the potential causes of containment failures have been identified leading to the study of possible countermeasures taking into account the combination of conservative design and post-TMI measures already implemented . A section of the paper will also be devoted to the specific study made for the selection, the sizing and the implementation of hydrogen control means. After the description of the selected measures implemented, the paper also describes the content of the 'Severe Accident Management Guidelines' developed by Tractebel for the Tihange NPPs and for the Doel NPPs. This project aimed at providing the operators with procedures or guidelines enabling to deal with complex situations not formally considered in the standard Emergency Response Guidelines, including accidents in which a significant portion of the core melts. The objective of these SAMG's programs is to indicate actions that must bring the plant to a controlled stable state and, above all, mitigate any challenges to the fission product barriers. The plant personnel must use the available plant information to determine the best severe accident management measures. Obviously

  11. Method for consequence calculations for severe accidents

    International Nuclear Information System (INIS)

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

  12. ALWR severe accident issue resolution in support of updated emergency planning

    International Nuclear Information System (INIS)

    The Advanced Light Water Reactor (ALWR) Program in the U.S. is a cooperative, cost-sharing undertaking between the U.S. government, industry, and a number of international participants, with the objective of developing the next generation of nuclear power plants. The ALWR designs emphasize improvements in safety and operational reliability through simplification, improved safety margins, innovative passive safety systems, enhanced man-machine interfaces, and incorporation of the lessons learned from the operation of existing LWR plants. An important component of the improved safety characteristics of ALWRs is the consideration of severe accidents in the plant design. The U.S. Department of Energy (DOE) initiated the Advanced Reactor Severe Accident Program (ARSAP) to assist in the transfer of severe accident technology from the U.S. national laboratories to the industry to implement this approach. The basic design requirements for this new generation of nuclear power plants were developed, under the management of the Electric Power Research Institute (EPRI) by the utilities and documented in the Utility Requirements Document (URD). The URD safety policy is based on the traditional 'defense-in-depth' approach, which emphasizes prevention through safety systems which prevent accidents from progressing to core damage, and mitigation to ensure that accidents are mitigated and contained. In a major departure from previous practice, severe accidents, including postulated core melt events, are specifically included in the defense-in-depth design considerations for ALWRs. As a result of this approach, the emergency planning assumptions and criteria warrant a review and reevaluation for ALWR designs. ALWRs present a risk profile that is significantly different than that which served as the basis for the emergency planning requirements for operating plants. The determination of this profile necessarily requires the characterization of the severe accident response of ALWRs

  13. A knowledge based severe accident handbook for PWR

    International Nuclear Information System (INIS)

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

  14. Optimizing severe accident containment filtered vent systems - Severe accident analysis and the dry and scrubber filter technology

    International Nuclear Information System (INIS)

    The accident at Fukushima has re-emphasized the importance of the capability to protect containment integrity during severe accidents. In the area of containment filtered vent systems, advances have been made since the original systems were installed in some plants in the late eighties and nineties. The paper describes new work in developing design specific requirements and system design.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    International Nuclear Information System (INIS)

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

  18. Modeling of BWR core meltdown accidents - for application in the MELRPI.MOD2 computer code

    International Nuclear Information System (INIS)

    This report summarizes improvements and modifications made in the MELRPI computer code. A major difference between this new, updated version of the code, called MELRPI.MOD2, and the one reported previously, concerns the inclusion of a model for the BWR emergency core cooling systems (ECCS). This model and its computer implementation, the ECCRPI subroutine, account for various emergency injection modes, for both intact and rubblized geometries. Other changes to MELRPI deal with an improved model for canister wall oxidation, rubble bed modeling, and numerical integration of system equations. A complete documentation of the entire MELRPI.MOD2 code is also given, including an input guide, list of subroutines, sample input/output and program listing

  19. Severities of transportation accidents involving large packages

    International Nuclear Information System (INIS)

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

  20. Severities of transportation accidents involving large packages

    Energy Technology Data Exchange (ETDEWEB)

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

    1978-05-01

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

  1. 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).

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

    International Nuclear Information System (INIS)

    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).

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    reporting traffic accidents. The second questionnaire was administered to stakeholders in the transportation field and was made to detect strengths, threats and opportunities for reporting traffic accidents within the police. This Ph.D. study contributes significantly to the literature about under......Under-reporting of traffic accidents is a well-discussed subject in traffic safety and it is well-known that the degree of under-reporting of traffic accidents is quite high in many countries. Nevertheless, very little literature has been made to investigate what causes the high degree of under......-reporting. The problem of under-reporting is not unique for traffic accidents as severe under-reporting is a challenge in many other fields of incident reporting. In other incidents fields with intended or unintended harm, research has investigated the behavioural reasons for why people choose to report an...

  4. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    Energy Technology Data Exchange (ETDEWEB)

    Cave, L.; Min, T.K.

    1980-04-01

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment.

  5. Determination of optimal LWR containment design, excluding accidents more severe than Class 8

    International Nuclear Information System (INIS)

    Information is presented concerning the restrictive effect of existing NRC requirements; definition of possible targets for containment; possible containment systems for LWR; optimization of containment design for class 3 through class 8 accidents (PWR); estimated costs of some possible containment arrangements for PWR relative to the standard dry containment system; estimated costs of BWR containment

  6. Structural evaluation of electrosleeved tubes under severe accident transients

    International Nuclear Information System (INIS)

    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

  7. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  11. Drug use and the severity of a traffic accident

    NARCIS (Netherlands)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

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

  13. Progress in LWR severe accident research at the Forschungszentrum Karlsruhe

    International Nuclear Information System (INIS)

    The R and D program at the Forschungszentrum Karlsruhe (FZK), performed within Project Nuclear Safety Research (PSF), is centered around phenomena and processes that could possibly endanger containment integrity of a large Pressurized Water Reactor after a severe accident. It includes activities on in-vessel accident progression, in-vessel steam explosion, hydrogen behavior and mitigation, ex-vessel melt behavior. The goal is to describe and quantify the governing mechanisms and to develop verified models and calculational tools which are able to predict maximum possible loads for realistic accident scenarios on full plant scale. (author)

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

    International Nuclear Information System (INIS)

    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 UO2–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

  15. Core structure heat-up and material relocation in a BWR short-term station blackout accident

    International Nuclear Information System (INIS)

    This paper presents an analytical and numerical analysis which evaluates the core-structure heat-up and subsequent relocation of molten core materials during a NWR short-term station blackout accident with ADS. A simplified one-dimensional approach coupled with bounding arguments is first presented to establish an estimate of the temperature differences within a BWR assembly at the point when structural material first begins to melt. This analysis leads to the conclusions that the control blade will be the first structure to melt and that at this point in time, overall temperature differences across the canister-blade region will not be more than 200 K. Next, a three-dimensional heat-transfer model of the canister-blade region within the core is presented that uses a diffusion approximation for the radiation heat transfer. This is compared to the one-dimensional analysis to establish its compatibility. Finally, the extension of the three-dimensional model to include melt relocation using a porous media type approximation is described. The results of this analysis suggest that under these conditions significant amounts of material will relocate to the core plate region and refreeze, potentially forming a significant blockage. The results also indicate that a large amount of lateral spreading of the melted blade and canister material into the fuel rod regions will occur during the melt progression process. 22 refs., 18 figs., 1 tab

  16. 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. PMID:23628941

  17. Severe Accident Management System On-line Network SAMSON

    International Nuclear Information System (INIS)

    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 cm2) in size to breaks 3.0 square feet in size (2800 cm2). (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  1. Severe accident countermeasure plan (draft) for nuclear power plants

    International Nuclear Information System (INIS)

    Nuclear power plants should be designed, constructed, and operated properly so that the likelihood of occurrence of a severe accident and its consequence may be minimized. The Korea Institute of Nuclear Safety has been reviewing the Nuclear Safety Policy Statement and the preceding severe accident countermeasure plan and prepared a new draft plan in order to provide a reasonable regulatory position for severe accidents. This plan has been prepared by taking into account the different reactor types and the characteristics of operating plants, new plants using the existing design, and new ones including the next generation plants. The major elements included in the plan are: establishment and application of the safety goal, performance of the probabilistic safety assessment and establishment of countermeasure plans for the vulnerabilities, provisions for severe accidents prevention and mitigation capability, set-up of a severe accident management program implementation system. Each element has been set up to move progressively toward an upgrading in safety of currently operating plants and future ones

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

    International Nuclear Information System (INIS)

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

  3. Sensitivity analysis in severe accidents semi-mechanistic modeling

    International Nuclear Information System (INIS)

    A sensitivity analysis to determine the most influent phenomena in the core melt progression to be considered in a semi-mechanistic modeling have been performed in the present work. The semi-mechanistic program MARCH3 and the TMI-2 plant parameters were used in the TMI-2 severe accident. The sensitivity analysis was performed with the comparison of the results obtained by the program with the plant data recorded during the accident. The results enabled us to verify that although many phenomena are present in the accident, the modelling of the most important ones was enough to reproduce, at least in a qualitative way, the accident progression. This fact reflects the importance of the sensitivity analysis to select the most influent phenomena in a core melting process. (author). 48 refs., 28 figs., 6 tabs

  4. Level 2 PRA for a German BWR

    International Nuclear Information System (INIS)

    A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2002-06-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Tusheva, Polina

    2012-03-09

    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

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

    International Nuclear Information System (INIS)

    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

  9. Evaluation of severe accident risks: Quantification of major input parameters

    International Nuclear Information System (INIS)

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported in NUREG-1150, the Severe Accident Risk Reduction Program (SAARP) has completed a revised calculation of the risk to the general public from severe accidents at five nuclear power plants: Surry, Sequoyah, Zion, Peach Bottom and Grand Gulf. The emphasis in this risk analysis was not on determining a point estimate of risk, but to determine the distribution of risk, and to assess the uncertainties that account for the breadth of this distribution. Off-site risk initiation by events, both internal to the power station and external to the power station. Much of this important input to the logic models was generated by expert panels. This document presents the distributions and the rationale supporting the distributions for the questions posed to the Source Term Panel

  10. Evaluation of severe accident risks: Quantification of major input parameters

    International Nuclear Information System (INIS)

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported in NUREG-1150, the Severe Accident Risk Reduction Program (SAARP) has completed a revised calculation of the risk to the general public from severe accidents at five nuclear power plants: Surry, Sequoyah, Zion, Peach Bottom, and Grand Gulf. The emphasis in this risk analysis was not on determining a ''so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiation by events, both internal to the power station and external to the power station were assessed. Much of the important input to the logic models was generated by expert panels. This document presents the distributions and the rationale supporting the distributions for the questions posed to the Structural Response Panel

  11. On the graphic simulation of severe accidents using IBM PC

    International Nuclear Information System (INIS)

    In this paper some methods of developing a graphical simulation system using IBM PC and MELCOR code were discussed. Most TH codes, which can calculate the severe accident phenomena parameters, are now running on workstations or supercomputers, to meet the complicated calculations. But the recent progress in computer engineering including personal computers make it possible to run TH codes in small microprocessor environments. Therefore we have surveyed some technical aspects in developing a severe accident graphical simulation system using IBM PC and MELCOR code. Especially noticing the complicated calculation results and the difficulties in obtaining continuous data due to the large data size, which are main defects of TH codes, we also surveyed the graphical methods to overcome these problems. The proposed methods are expected to play important roles in developing our graphic simulation system of severe accidents

  12. Plant specific severe accident management - the implementation phase

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    1989-07-01

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

  14. Investigation of VVER 1000 Fuel Behavior in Severe Accident Condition

    International Nuclear Information System (INIS)

    This paper presents the results obtained during a simulation of fuel behavior with the MELCOR computer code in case of severe accident for the VVER reactor core. The work is focused on investigating the influence of some important parameters, such as porosity, on fuel behavior starting from oxidation of the fuel cladding, fusion product release in the primary circuit after rupture of the fuel cladding, melting of the fuel and reactor core internals and its further relocation to the bottom of the reactor vessel. In the analyses are modeled options for blockage of melt and debris during its relocation. In the work is investigated the uncertainty margin of reactor vessel failure based on modeling of the reactor core and an investigation of its behavior. This is achieved by performing sensitivity analyses for VVER 1000 reactor core with gadolinium fuel type. The paper presents part of the work performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) in the frame of severe accident research. The performed work continues the effort in the modeling of fuel behavior during severe accidents such as Station Blackout sequence for VVER 1000 reactors based on parametric study. The work is oriented towards the investigation of fuel behavior during severe accident conditions starting from the initial phase of fuel damaging through melting and relocation of fuel elements and reactor internals until the late in-vessel phase, when melt and debris are relocated almost entirely on the bottom head of the reactor vessel. The received results can be used in support of PSA2 as well as in support of analytical validation of Sever Accident Management Guidance for VVER 1000 reactors. The main objectives of this work area better understanding of fuel behavior during severe accident conditions as well as plant response in such situations. (author)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-07-01

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

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

    International Nuclear Information System (INIS)

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

  18. Nuclear safety in light water reactors severe accident phenomenology

    CERN Document Server

    Sehgal, Bal Raj

    2011-01-01

    This vital reference is the only one-stop resource on how to assess, prevent, and manage severe nuclear accidents in the light water reactors (LWRs) that pose the most risk to the public. LWRs are the predominant nuclear reactor in use around the world today, and they will continue to be the most frequently utilized in the near future. Therefore, accurate determination of the safety issues associated with such reactors is central to a consideration of the risks and benefits of nuclear power. This book emphasizes the prevention and management of severe accidents to teach nuclear professionals

  19. Severe accident considerations in Canadian nuclear power reactors

    International Nuclear Information System (INIS)

    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

  20. Formulating the Canadian regulatory position on severe accidents

    International Nuclear Information System (INIS)

    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

  1. MAAP4.0.7 severe accident source term analysis

    International Nuclear Information System (INIS)

    The Severe Accident Source Term Analysis performed in support of U.S. EPR design certification was conducted using MAAP4.07. The analysis had three distinct goals: to determine the most limiting scenario from a severe accident stand point and incorporate the annulus, fuel and safeguards buildings into the MAAP4.0.7 base model; to develop and document the Level 2 Probabilistic Risk Assessment (PRA) Source Term Analysis; and to develop the input from the PRA Level 2 output to PRA Level 3. The methods of this analysis will be presented in this paper. (authors)

  2. Definition of loss-of-coolant accident radiation source. [PWR; BWR

    Energy Technology Data Exchange (ETDEWEB)

    1978-02-01

    Meaningful qualification testing of nuclear reactor components requires a knowledge of the radiation fields expected in a loss-of-coolant accident (LOCA). The overall objective of this program is to define the LOCA source terms and compare these with the output of various simulators employed for radiation qualification testing. The basis for comparison will be the energy deposition in a model reactor component. The results of the calculations are presented and some interpretation of the results given. The energy release rates and spectra were validated by comparison with other calculations using different codes since experimental data appropriate to these calculations do not exist.

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

    International Nuclear Information System (INIS)

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

  4. Studies of severe accidents in light-water reactors

    International Nuclear Information System (INIS)

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

  5. Spent Fuel Pool Decommissioning After a Severe Accident. Appendix

    International Nuclear Information System (INIS)

    Most decommissioning related publications by the IAEA [A.1–A.4] and other organizations clearly specify that their scope applies to the decommissioning of nuclear facilities under planned conditions. It is generally specified that decommissioning of facilities that have been subject to a severe accident is excluded from the scope of these publications. This is because of the peculiar, and generally unpredictable, circumstances resulting from a severe accident, including, among others, high radiation and contamination fields, abnormal waste and unexpected configuration changes. Based on the literature, there is no unique definition of a severe accident. All definitions include various consequence (damage) types (evacuees, injured persons, fatalities or costs) and a minimum level for each damage type. The differences between the definitions concern both the set of specific consequence types considered and the damage threshold. For the purposes of this publication, the scope of this Appendix encompasses only facilities (spent fuel pools) that have been seriously contaminated and physically damaged to the point that planned routine decommissioning strategies and techniques are unusable or impractical. It should be noted that there are three phases typically associated with a post-accident phase: stabilization, recovery and decommissioning. Stabilization refers to the immediate aftermath of a nuclear accident, and implies controlling of conditions so that impacts to the environment and public are controlled and minimized. Recovery entails the planning and implementation of activities to limit, and subsequently reduce, the extent of abnormal conditions, and prepare the plant for achievement of a longer term, safer configuration. Recovery can be viewed as a precursor to decommissioning. However, there is no clear-cut line between the three above mentioned phases. In fact, conditions generated by the accident and its evolution may initially be recognized, faced and dealt

  6. Iodine–paint interactions during nuclear reactor severe accidents

    International Nuclear Information System (INIS)

    Highlights: • Iodine interacts with containment paint in several ways. • Some mechanistic understanding is required. • Paint aging and degradation mechanisms. • Iodine adsorption and release mechanisms. • The Severe Accident Research Network (SARNET) facilitates collaboration between members. - Abstract: To assess the radiological consequences of a severe reactor accident, it is important to be able to predict the behaviour of iodine in containment. Some interactions between iodine and containment paint (e.g., adsorption) have been well known for a long time. However, in recent years, new phenomena have been identified that can affect the gas phase iodine concentration in the longer term (e.g., the release of molecular iodine and organic iodides from irradiated painted surfaces). Several international collaborations and organizations around the world are currently addressing different aspects of this topic, including laboratory experiments and theoretical studies (ab initio) designed to improve the mechanistic understanding of the phenomena. Knowledge of the underlying mechanisms will provide explanations for behavioural differences observed between paint types, and will support the extrapolation of laboratory results to the safety analyses of nuclear reactors. The purpose of this paper is to present a selection of recent work performed by Severe Accident Research Network (SARNET) members regarding iodine–paint interactions and paint aging in order to improve the common understanding and better define what has still to be done in this area. The Severe Accident Research Network (SARNET) provides a framework within which members can share and discuss results

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

    International Nuclear Information System (INIS)

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

  8. IDCOR: the technical foundation and process for severe accident decisions

    International Nuclear Information System (INIS)

    The Industry Degraded Core Rulemaking (IDCOR) Program has been successful in establishing a technical foundation for pursuing the resolution of severe accident issues. IDCOR is supported by the 62 nuclear utilities, architect-engineers, and light water reactor (LWR) vendors in the United States and by Japan and Sweden. The IDCOR mission was to develop a comprehensive, technically sound position on the issues related to potential severe accidents in nuclear power plant light water reactors. An intensive two and one-half year technical program was completed on schedule and within budget in July 1983. IDCOR identified key issues and phenomena; developed analytical methods; analyzed the severe accident behaviour of four representative plants; and extended the results as generically as possible. In general, IDCOR has demonstrated that consequences of dominant accident sequences are significantly less than previously anticipated. Most accident sequences require long times to progress, allowing time to achieve safe stable states. In July 1983, IDCOR entered a second phase to complete industry and expert review of IDCOR results, perform a few additional technical evaluations, publish the work, and explain the results to the technical and regulatory community. IDCOR results are contained in technical reports capped by a substantial main technical summary report and a small results and conclusions report. Some final reports have been made available to the technical community and to the Nuclear Regulatory Commission (NRC). All reports presently are being finalized and printed and will be made available in a logical sequence of meetings with the NRC over the next several months. Concurrent with IDCOR efforts, the NRC has published a policy paper and a proposed decision-making process for reaching permanent resolution of the severe accident issues

  9. The assessment of the Fukushima like accident for Kuosheng BWR/6 with TRACE

    International Nuclear Information System (INIS)

    After the Fukushima Daiichi nuclear disaster, the Taiwan Power Company has developed a procedure to be used for Fukushima like accidents and is named URG (Ultimate Response Guideline). The main content of the URG is a DIVing plan, which abbreviates from two step depressurization, cooling water injection, and containment venting. It is claimed that with the DIVing plan, the cladding temperature could be kept under 1088.6K (1500 °F) such that no hydrogen would be generated and the core could be prevented from melting. In this study, the Fukushima like accident for the Kuosheng nuclear power station was simulated by TRACE (TRAC/RELAP Advanced Computational Engine), which is an advanced thermal hydraulics code and has the capability of modeling thermal hydraulic phenomena in both 1-D and 3-D space. Two scenarios were considered to verify the effectiveness of the DIVing plan. The results show that the reactor will be brought to and maintained at a relatively safe state with the slow and controlled depressurization process in the first step, and then the impact of the emergency depressurization following the failure of RCIC will be minimized. On the contrary, the fast depressurization will lead to a core uncovered state without the controlled depressurization process. Besides, the effectiveness associated with the controlled depressurization and the emergency depressurization was verified through the investigation of the number of safety relief valves opened in the sensitivity study. (author)

  10. LWR severe accident source term research in the USA

    International Nuclear Information System (INIS)

    Fission product releases to the environment, or source terms, arise as a result of a highly diverse group of phenomena involved in any particular severe accident sequence. Because of the multiplicity of accident sequences that can occur for a given plant as well as the diversity of the, as yet, imperfectly understood severe accident phenomena, it is not surprising that reactor accidents such as, for example, those documented in NUREG-1150 have indicated large uncertainties in source terms which represent a significant contribution to the uncertainty in the absolute value of risk. Because of the difficulty and expense involved in performing prototypic experiments, substantial reliance has been placed on the development and validation of detailed mechanistic computer codes for analyzing severe accident phenomena and the source terms associated with them. This paper discusses the extensive research and other efforts that have taken place over the last decade to address the technical issues which have a bearing on being able to describe quantitatively the source term(s) and its characteristics. It also summarizes our present state of knowledge and points out areas where additional research will add further to our understanding. In this context the paper discusses the information that could be provided by the PHEBUS-FP program and its use to assess severe accident integral evaluation codes such as VICTORIA and CONTAIN. Finally, this paper discusses the United States Nuclear Regulatory Commission 's efforts to revise the licensing source term (TID-14844) and the implications of this revision, especially for siting and design of future power plants. (author)

  11. Iodine chemistry and associated interactions under severe accident conditions

    International Nuclear Information System (INIS)

    In a highly improbable severe accident wherein the core cooling is decapacitated or insufficient the scenario may lead to melting of fuel elements and fission products release. Nuclear power plants are designed with inherent engineering safety systems and associated operational procedures that provide an in-depth defence against such accidents. Iodine, one of the fission products, behaviour is required for the analysis of severe accident consequences because iodine is a chemical more active to the potential source term for release 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-bearing particles is governed by aerosol physics, depletion mechanisms gravitational settling, diffusiophoresis and thermophoresis. Sorption and desorption of iodine occurring on containment surface are also of importance. 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 plays significant roles. Water radiolysis products due to presence of dissolved impurities such as dissolved oxygen, nitrate/nitrite (NO3/NO2) produced by air radiolysis, trace metal ions such as Fe2+/Fe3+ dissolved from steel surfaces, chloride ions coming from the pyrolysis/radiolysis of polyvinyl material from cables and organic impurities from painted surfaces and polymers also inherent and should be considered while calculating iodine release. This paper elaborates stare of art on iodine chemistry and its behaviour during accident. (author)

  12. ALMOS-2 - computer program for accident analyses of BWR-type reactor plants. Pt. 1

    International Nuclear Information System (INIS)

    ALMOS-2 is a non-linear, one-dimensional digital computer program for accident analyses of boiling water reactor plants. The program is capable of calculating transients with large amplitude deviations from steady state conditions, such as pump failures, loss of heat sink, feedwater system failures, control rod malfunction and safety system failures. The program is a suitable tool for the analysis of Anticipated Transients Without Scram (ATWS). The model includes the reactor core, the main coolant system and the safety system. Boundary conditions to these systems can be defined as functions of time by input data. The program description includes a short introduction into the physical model, a presentation of the program structure and program components and a detailed input data description. For test purposes the results of a transient calculation are added. (orig.)

  13. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway. 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. A second objective of the program was to develop an interactive graphical user interface which presented the system's information in an easily accessible and straightforward manner to the user. This paper describes the technical and regulatory foundation upon which the expert system is based and provides a background on the development of a new severe accident management tool. This tool provides data to assist in; (1) planning and developing priorities for recovery actions, (2) evaluating recovery action feasibility, (3) identifying recovery action options, and (4) assessing the timing and possible effects of potential recovery strategies. These performance characteristics represent the goals identified for the Severe Accident Management Strategies Online Network (SAMSON) which is currently under development. 4 refs, 1 fig., 1 tab

  15. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  16. The development of a severe accident analysis code

    International Nuclear Information System (INIS)

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

  17. Code assessment in context of severe accident phenomenology

    Energy Technology Data Exchange (ETDEWEB)

    Bratfisch, C.; Agethen, K.; Braehler, T.; Risken, T.; Koppers, V.; Gremme, F.; Hoffmann, M.; Koch, M.K. [Bochum Univ. (Germany). Reactor Simulation and Safety Group

    2014-05-15

    The following paper gives an outline of current research activities in the field of reactor simulation and safety at Ruhr-Universitaet Bochum. Results related to phenomena of core degradation, hydrogen combustion and molten corium concrete interaction will be presented. These deal with the simulation of relevant experiments in order to validate the severe accident codes ASTEC, ATHLET-CD and COCOSYS. Exemplarily, simulation results of the tests QUENCH-16, BMC Ix9 and OECD CCI-2/-3 are discussed. The importance of these phenomena is illustrated by the Three Mile Island and Fukushima Daiichi accidents. (orig.)

  18. Severe accident analysis code Sampson for impact project

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  19. Severe accident analysis code Sampson for impact project

    International Nuclear Information System (INIS)

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

  20. Severe accident natural circulation studies at the INEL

    International Nuclear Information System (INIS)

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

  1. Test Data for USEPR Severe Accident Code Validation

    Energy Technology Data Exchange (ETDEWEB)

    J. L. Rempe

    2007-05-01

    This document identifies data that can be used for assessing various models embodied in severe accident analysis codes. Phenomena considered in this document, which were limited to those anticipated to be of interest in assessing severe accidents in the USEPR developed by AREVA, include: • Fuel Heatup and Melt Progression • Reactor Coolant System (RCS) Thermal Hydraulics • In-Vessel Molten Pool Formation and Heat Transfer • Fuel/Coolant Interactions during Relocation • Debris Heat Loads to the Vessel • Vessel Failure • Molten Core Concrete Interaction (MCCI) and Reactor Cavity Plug Failure • Melt Spreading and Coolability • Hydrogen Control Each section of this report discusses one phenomenon of interest to the USEPR. Within each section, an effort is made to describe the phenomenon and identify what data are available modeling it. As noted in this document, models in US accident analysis codes (MAAP, MELCOR, and SCDAP/RELAP5) differ. Where possible, this report identifies previous assessments that illustrate the impact of modeling differences on predicting various phenomena. Finally, recommendations regarding the status of data available for modeling USEPR severe accident phenomena are summarized.

  2. ANS severe accident program overview ampersand planning document

    International Nuclear Information System (INIS)

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

  3. Evaluation of severe accident risks, Peach Bottom, Unit 2: Appendices

    International Nuclear Information System (INIS)

    In support of the Nuclear Regulatory Commission's (NRC's) assessment of the risk from severe accidents at commercial nuclear power plants in the US reported in NUREG-1150, the Severe Accident Risk Reduction Program (SARRP) has completed a revised calculation of the risk to the general public from severe accidents at the Peach Bottom Atomic Power Station, Unit 2. This power plant, located in southeastern Pennsylvania, is operated by the Philadelphia Electric Company. The emphasis in this risk analysis was not on determining a ''so-called'' point estimate of risk. Rather, it was to determine the distribution of risk, and to discover the uncertainties that account for the breadth of this distribution. Off-site risk initiated by events both internal and external to the power station were assessed. This document provides Appendices A through E which include the following topics respectively: accident progression event tree; supporting information for the source term analysis; supporting information for the consequence analysis; risk results; and sampling information. 6 figs., 6 tabs

  4. Severe accident natural circulation studies at the INEL

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-02-01

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

  5. Aerosol Characterization in Containment Air during Severe Accident

    International Nuclear Information System (INIS)

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

  6. Finite element based stress analysis of BWR internals exposed to accident loads

    Energy Technology Data Exchange (ETDEWEB)

    Altstadt, E.; Weiss, F.P.; Werner, M.; Willschuetz, H.G.

    1998-10-01

    During a hypothetical accident the reactor pressure vessel internals of boiling water reactors can be exposed to considerable loads resulting from temperature gradients and pressure waves. Three dimensional FE models were developed for the core shroud, the upper and the lower core supporting structure, the steam separator pipes and the feed water distributor. The models of core shroud, upper core structure and lower core structure were coupled by means of the substructure technique. All FE models can be used for thermal and for structural mechanical analyses. As an example the FE analysis for the case of a station black-out scenario (loss of power supply for the main circulating pumps) with subsequent emergency core cooling is demonstrated. The transient temperature distributions within the core shroud and within the steam dryer pipes as well were calculated based on the fluid temperatures and the heat transfer coefficients provided by thermo-hydraulic codes. At the maximum temperature gradients in the core shroud, the mechanical stress distribution was computed in a static analysis with the actual temperature field being the load. (orig.)

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

    International Nuclear Information System (INIS)

    Comparative assessment of accident risks in the energy sector is a key aspect in a comprehensive evaluation of sustainability and energy security concerns. Safety performance of energy systems can have important implications on the environmental, economic and social dimensions of sustainability as well as availability, acceptability and accessibility aspects of energy security. Therefore, this study provides a broad comparison of energy technologies based on the objective expression of accident risks for complete energy chains. For fossil chains and hydropower the extensive historical experience available in PSI's Energy-related Severe Accident Database (ENSAD) is used, whereas for nuclear a simplified probabilistic safety assessment (PSA) is applied, and evaluations of new renewables are based on a combination of available data, modeling, and expert judgment. Generally, OECD and EU 27 countries perform better than non-OECD. Fatality rates are lowest for Western hydropower and nuclear as well as for new renewables. In contrast, maximum consequences can be by far highest for nuclear and hydro, intermediate for fossil, and very small for new renewables, which are less prone to severe accidents. Centralized, low-carbon technology options could generally contribute to achieve large reductions in CO2-emissions; however, the principal challenge for both fossil with Carbon Capture and Storage and nuclear is public acceptance. Although, external costs of severe accidents are significantly smaller than those caused by air pollution, accidents can have disastrous and long-term impacts. Overall, no technology performs best or worst in all respects, thus tradeoffs and priorities are needed to balance the conflicting objectives such as energy security, sustainability and risk aversion to support rationale decision making. - Highlights: • Accident risks are compared across a broad range of energy technologies. • Analysis of historical experience was based on the

  8. Some outstanding issues in severe accidents containment performance

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Martin, Robert P.

    2012-01-01

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

  10. Experimental data report for test TS-3 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    This report presents experimental data for Test TS-3 which was the third test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in September, 1990. Test fuel rod used in the Test TS-3 was a short-sized BWR (7 x 7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79 % and a burnup of 26 Gwd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The energy deposition of the fuel rod in this test was evaluated to be 94 ± 4 cal/g · fuel (88 ± 4 cal/g · fuel in peak fuel enthalpy) and no fuel failure was observed. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

  11. Experimental data report for test TS-5 Reactivity Initiated Accident test in the NSRR with pre-irradiated BWR fuel rod

    International Nuclear Information System (INIS)

    This report presents experimental data for Test TS-5 which was the fifth test in a series of Reactivity Initiated Accident (RIA) tests using pre-irradiated BWR fuel rods, performed in the Nuclear Safety Research Reactor (NSRR) in January, 1993. Test fuel rod used in the Test TS-5 was a short-sized BWR (7x7) type rod which was re-fabricated from a commercial rod irradiated in the Tsuruga Unit 1 power reactor of Japan Atomic Power Co. The fuel had an initial enrichment of 2.79% and a burnup of 26GWd/tU. A pulse irradiation of the test fuel rod was performed under a cooling condition of stagnant water at atmospheric pressure and at ambient temperature which simulated a BWR's cold start-up RIA event. The nominal energy deposition of 117±5cal/g·fuel (98±4cal/g·fuel in peak fuel enthalpy) was subjected to the test fuel rod and no fuel failure was observed in the test. The test fuel was pulse irradiated in a flow shroud which simulates fuel/water ratio in the commercial assembly. Descriptions on test conditions, test procedures, transient behavior of the test rod during the pulse irradiation, and results of pre-pulse and post-pulse irradiation examinations are described in this report. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  14. Trends in BWR transient analysis

    International Nuclear Information System (INIS)

    While boiling water reactor (BWR) analysis methods for transient and loss of coolant accident analysis are well established, refinements and improvements continue to be made. This evolution of BWR analysis methods is driven by the new applications. This paper discusses some examples of these trends, specifically, time domain stability analysis and analysis of the simplified BWR (SBWR), General Electric's design approach involving a shift from active to passive safety systems and the elimination/simplification of systems for improved operation and maintenance

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Estimated consequences from severe spent nuclear fuel transportation accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. Development and validation of Maanshan severe accident management guidelines

    International Nuclear Information System (INIS)

    Maanshan is a Westinghouse pressurized water reactor Nuclear Power Plant (NPP) located in south Taiwan. The Severe Accident Management Guideline (SAMG) of Maanshan NPP is developed based on the Westinghouse Owners Group (WOG) SAMG. The Maanshan SAMG is developed at the end of 2002. MAAP4 code is used as tool to validate the SAMG strategies. The development process and characteristics of Maanshan SAMG is described. A Station BlackOut (SBO) accident for Maanshan NPP which occurred in March 2001 is cited as a reference case for SAMG validation. A SBO accident is simulated first. The severe accident progression is simulated and the entry condition of SAMG is described. Mitigation actions are then applied to demonstrate the effect of SAMG. A RCS depressurization, RCS injection, and containment hydrogen reduction strategies are used to restore the system to a stable condition as power is recovered. Hot leg creep rupture is occurs during the mitigation action that is not considered in WOG SAMG. The effect of the RCS depressurization, RCS injection, and containment hydrogen reduction strategies are analyzed with MAAP4 code

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

    International Nuclear Information System (INIS)

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

  1. Hydrogen distribution during postulated severer accident in kaiga containment

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-07-01

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

  2. Hydrogen distribution during postulated severer accident in kaiga containment

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1987-02-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  6. Numerical module for debris behavior under severe accident conditions

    International Nuclear Information System (INIS)

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

  7. Overview of severe accident research at the USNRC

    International Nuclear Information System (INIS)

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

  8. Behaviour of fission-product iodine under severe accident conditions

    International Nuclear Information System (INIS)

    On account of the radiological properties of I-131 the behaviour of fission-product iodine is of great importance under severe reactor accident conditions. The chemical properties of iodine: Its easy conversion into several oxidation compounds, its capability of forming not only volatile (organo-iodide, elemental iodine), hardly volatile, readily soluble (cesium iodide/iodate) but also insoluble (silver iodide) compounds, and its susceptibility to ionizing radiation, are further aspects of significance. Intensive investigations on iodine behaviour under reactor accident conditions carried out worldwide over the last ten years have shown - even though a number of details have yet to be elucidated - that physicochemical processes form a natural, i.e. passive, barrier against the possible release of iodine. (orig.)

  9. Analysis of progression of severe accident in Indian PHWRs

    International Nuclear Information System (INIS)

    In India a wide variety of nuclear reactors are in operation and in different stages of construction. The main stay of Indian nuclear power programme today is 'Pressurised Heavy Water Reactors (PHWRs)'. There are 13 operating PHWRs and several others in different stages of construction. These reactors are either of 220 MWe or 540 MWe capacity. Atomic Energy Regulatory Board for authorization needs safety analysis reports, which consists of a detailed analyses of all design basis accidents. However, there is a more to carry out severe accident analysis for accident management programme. This paper describes an analysis of a severe accident caused by Loss of Coolant Accident (LOCA), co-incident with loss of emergency core cooling system and loss of moderator heat sink in 220 MWe Indian PHWR. Initially in a matter of about 60 seconds most of the coolant from primary heat transport system blows out, the reactor gets tripped, but in the absence of emergency core cooling system, the heat removal from the fuel bundles is very poor. Consequently, the fuel bundle starts getting heated up. The only mode of heat transfer is radiative heat transfer from fuel bundle to pressure tube, from pressure tube to calandria tube and them convective heat transfer from calandria tube to the relatively cold moderator in which the reactor channels are immersed. In the absence of availability of moderator heat sink the moderator gets heated up and eventually boils. And slowly the moderator level in the calandria starts falling. Soon the channel gets uncovered and the temperatures of the channel components shoot up as the temperature of the pressure tube and calandria tube rise. Mechanical properties deteriorate rapidly with temperature as structural elements of reactor channel are made of zircaloy. Under the weight of the fuel, the reactor channel gives in and falls into the remaining moderator. This process continues till all the moderator is evaporated, leading to damage to the entire

  10. Dependence of severe accident source terms on containment performance

    International Nuclear Information System (INIS)

    The results of the BMI-2104 analyses for the assessment of severe accident fission product source terms from the point of view of the effectiveness of the containment in attenuating the radioactivity that is predicted to be released to it are examined. The containment performance assumptions considered include design leakage, leak-before-break, large threshold failure and failure to isolate. The results are expressed and discussed in terms of the effective containment decontamination factors inferred from these analyses

  11. Experimental investigations on nuclear aerosols in a severe accident

    OpenAIRE

    DELGADO TARDÁGUILA, ROSARIO

    2016-01-01

    [EN] In case of a severe accident in a NPP fission products are released from the degraded fuel and may reach the environment if their confinement is lost and/or bypassed. Given the high radio-toxic nature of nuclear aerosols for environment and population, their unrestricted release should be absolutely avoided. One particular situation is the core meltdown sequence with steam generator tube rupture (SGTR). The containment bypass turns this sequence into an indispensable scenario to mode...

  12. Research Needs in the Domain of Severe Accidents

    International Nuclear Information System (INIS)

    The objectives of SARNET are to define common research programmes and to develop common computer tools and methodologies for safety assessment. To reach these objectives several elements or work programmes (WP) are established. One of them is the WP 'Severe Accident Research Priorities' (SARP) with the aims to harmonize and to re-orient research programmes, to define new ones, and to close a resolved issue on a common basis. This action will make use notably of (1) the outcome of the EURSAFE action, i.e., the results of the Phenomena Identification and Ranking Tables (PIRT) on severe accidents, (2) the results of the qualification and benchmarking activities on ASTEC, (3) the results of reactor calculations carried out in the other activities, and (4) the outcome of the research performed in the three thematic sub domains of SARNET (corium, containment, source term). The main outcome of EURSAFE was a list of 21 topics which includes precise recommendations for experimental programmes and code developments and forms the basis of the work of the SARP. Also the methodology applied in EURSAFE to consider both risk potential and the severe accident issues where large uncertainties still subsist will be adopted. The analyses of the progress of research and developmental activities will be in close cooperation with the management team and the coordinators of the WPs. These analyses will consider whether (1) any research issue is resolved due to reduction of uncertainties or gain of scientific insights, (2) any new issue has to be added to the list of needed research, (3) any new process or phenomena have to be included in the general PIRT list taking into account the safety relevance and lack of knowledge, and (4) a new accident management programme has to be developed to cope with unresolved problems. Furthermore a strategy plan will be elaborated to ensure a wide consensus with the end users requirements and the objectives of SARNET research activities. (author)

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

    International Nuclear Information System (INIS)

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

  14. Regulatory analyses for severe accident issues: an example

    International Nuclear Information System (INIS)

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

  15. Westinghouse severe accident management guidance overview and current status

    International Nuclear Information System (INIS)

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

  16. Definition of loss-of-coolant accident radiation source: summary and conclusions. [BWR; PWR

    Energy Technology Data Exchange (ETDEWEB)

    Bonzon, L.L.; Lurie, N.A.; Houston, D.H.; Naber, J.A.

    1978-05-01

    The radiation energy release rates and spectra corresponding to those sources specified in USNRC Regulatory Guide 1.89 for the radiation qualification of Class 1E equipment were calculated. The effects of several parameters (some not specific in the Guide), such as reactor fuel composition, operating duration and power level, and treatment of progeny, are evaluated. The results are presented as time-dependent beta and gamma-ray energy release rates and spectra which are fundamental quantities that are not specific to a plant design but are generally applicable to any nuclear power station.

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

    OpenAIRE

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

    2010-01-01

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

  18. Radionuclide release calculations for selected severe accident scenarios

    International Nuclear Information System (INIS)

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

  19. Radionuclide release calculations for selected severe accident scenarios

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-08-01

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

  20. Containment in-situ post accident sampling and emission monitoring for design basis and severe accidents

    International Nuclear Information System (INIS)

    In order to reduce the residual risk in case of severe accident the German Reactor Safety Commission imposed additional recommendations that provision are made for: Post Accident Sampling System (PASS ) for the confinement atmosphere for obtaining of information on the condition of the core and potential hazards to the environment; Emission monitoring for containment venting. In-situ PASS provides the following information: activity concentration of radioactive aerosols, gaseous iodine and noble gases and composition of gases present in the containment atmosphere (CO, CO2, O2 etc.).The emission monitoring system has the task of monitoring radioactive materials discharged during DBA or severe accident if filtered confinement venting will be performed. The data obtained from the monitoring serves as a base for the implementing of the emergency response actions. The system is designed for the maximum activity concentration at the time of early containment venting, commenced in the event of a core melt accident - 1011 Bq/m3 aerosol bound radionuclides; 5.1011 Bq/m3 gaseous iodine isotopes and 2.1015 Bq/m3 radioactive noble gases

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

    International Nuclear Information System (INIS)

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

  2. EPR design features to mitigate severe accident challenges

    International Nuclear Information System (INIS)

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

  3. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-15

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

  4. Level 2 PSA methodology and severe accident management

    International Nuclear Information System (INIS)

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

  5. Severe accident analysis code SAMPSON improvement for IMPACT project

    International Nuclear Information System (INIS)

    SAMPSON is the integral code for severe accident analysis in detail with modular structure, developed in the IMPACT project. Each module can run independently and communications with multiple analysis modules supervised by the analysis control module makes an integral analysis possible. At the end of Phase 1 (1994-1997), demonstration simulation by combinations of up to 11 analysis modules had been performed and physical models in the code had been verified by separate-effect tests and validated by integral tests. Multi-dimensional mechanistic models and theoretical-based conservation equations have been applied, during Phase 2 (1998 - 2000). New models for Accident Management evaluation have been also developed. Verification and validation have been performed by analysing separate-effect tests and integral tests, while actual plant analyses are also being in progress. (author)

  6. Long operating cycle simplified BWR

    International Nuclear Information System (INIS)

    Considering next generation requirement for nuclear plants, a long cycle operating simplified BWR (LSBWR) concept is proposed. The major features of LSBWR are; 1) Long cycle operation core using uranium fuels; 2) Simplified system and component as well as passive systems; 3) Combined building concept with ship hull structure. This concept have potential to reduce construction cost and to Increase availability. Safety feature of LSBWR makes possible to attain no evacuation capability in case of a severe accident. Further research and development is underway. (author)

  7. Fuel behavior in severe accidents and Mo-alloy based cladding designs to improve accident tolerance

    International Nuclear Information System (INIS)

    The severe accidents at TMI-2 and Fukushima-Daiichi led to core meltdown and hydrogen explosions. The main source of energy causing core melting is the decay heat from β-, β+, and γ decays of short-lived isotopes following a power scram. The exothermic reaction of Zr-alloy cladding can further increase the cladding temperature leading to rapid cladding corrosion and hydrogen production. The most effective mitigation to minimize core damage in a severe accident is to extend the duration of heat removal capacity via battery-supported passive cooling for as long as practically possible. Replacing the Zr-alloy cladding with a higher heat resistant cladding with lower enthalpy release rate may also provide additional coping time for accident management. Such a heat resistant cladding may also overcome the current licensing concerns about Zr-alloy hydriding and post quench ductility issues in a design base loss of coolant accident (LOCA). Zr-alloy cladding, while has been optimized for normal operation in high pressure water and steam of light water reactors, will rapidly lose its corrosion resistance and tensile and creep strength in high pressure steam. Evaluation of alternate cladding materials and designs have been performed to search for a new fuel cladding design which will substantially improve the safety margins at elevated temperatures during a severe accident, while maintaining the excellent fuel performance attributes of the current Zr-alloy cladding. The screening criteria for the evaluation include neutronic properties, material availability, adaptability and operability in current LWRs, resistance to melting. The new designs also need to be fabricable, maintain sufficient strength and resist to attack by high pressure steam. Engineering metals, alloys and ceramics which can meet some or most of these requirements are limited. Following review of the properties of potential candidates, it is concluded that molybdenum alloys may potentially achieve the

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

    International Nuclear Information System (INIS)

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

  9. Effect of Candu Fuel Bundle Modeling on Sever Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Dupleac, D.; Prisecaru, I. [Power Plant Engineering Faculty, Politehnica University, 313 Splaiul Independentei, 060042, sect. 6, Bucharest (Romania); Mladin, M. [Institute for Nuclear Research, Pitesti-Mioveni, 115400 (Romania)

    2009-06-15

    In a Candu 6 nuclear power reactor fuel bundles are located in horizontal Zircaloy pressure tubes through which the heavy-water coolant flows. Each pressure tube is surrounded by a concentric calandria tube. Outside the calandria tubes is the heavy-water moderator contained in the calandria itself. The moderator is maintained at a temperature of 70 deg. C by a separate cooling circuit. The moderator surrounding the calandria tubes provides a potential heat sink following a loss of core heat removal. The calandria vessel is in turn contained within a shield tank (or reactor vault), which provides biological shielding during normal operation and maintenance. It is a large concrete tank filled with ordinary water. During normal operation, about 0.4% of the core's thermal output is deposited in the shield tank and end shields, through heat transfer from the calandria structure and fission heating. In a severe accident scenario, the shield tank could provide an external calandria vessel cooling which can be maintained until the shield tank water level drops below the debris level. The Candu system design has specific features which are important to severe accidents progression and requires selective consideration of models, methods and techniques of severe accident evaluation. Moreover, it should be noted that the mechanistic models for severe accident in Candu system are largely less well validated and as the result the level of uncertainty remains high in many instances. Unlike the light water reactors, for which are several developed computer codes to analyze severe accidents, for Candu severe accidents analysis two codes were developed: MAAP4-Candu and ISAAC. However, both codes started by using MAAP4/PWR as reference code and implemented Candu 6 specific models. Thus, these two codes had many common features. Recently, a joint project involving Romanian nuclear organizations and coordinated by Politehnica University of Bucharest has been started. The purpose

  10. ESFR Severe Accident Analyses with SIMMER-III

    International Nuclear Information System (INIS)

    The Collaborative Project on European Sodium Fast Reactor, CP-ESFR, combines European efforts advancing fast reactor technology towards economics, safety and nuclear waste reduction. A key issue of development is the promise of a higher and improved safety level. Both on the prevention and mitigation side significant efforts are invested to fulfill the high safety goals. Research in severe accident phenomenology and safety analyses help to develop means for better prevention and mitigation. Within this framework accident initiators are investigated leading to an unprotected loss-of-flow (ULOF) and a total instantaneous blockage (TIB) scenario. Simulations focusing on the energetics behavior apply SIMMER-III, an advanced accident code coupled with space- and energy-dependent neutronics. For the ULOF especially the transition phase with its recriticality potential has been of interest, while for the TIB the issue of melt propagation has been a key focus. In addition it has been investigated whether the available core material removal paths are sufficiently effective to prevent recriticality scenarios. The ULOF conditions for SIMMER have been provided by a SAS-SFR simulation of the ULOF initiation phase. For the TIB the SIMMER simulations started from steady state core conditions. (author)

  11. Markov Model of Severe Accident Progression and Management

    Energy Technology Data Exchange (ETDEWEB)

    Bari, R.A.; Cheng, L.; Cuadra,A.; Ginsberg,T.; Lehner,J.; Martinez-Guridi,G.; Mubayi,V.; Pratt,W.T.; Yue, M.

    2012-06-25

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

  12. Markov Model of Severe Accident Progression and Management

    International Nuclear Information System (INIS)

    The earthquake and tsunami that hit the nuclear power plants at the Fukushima Daiichi site in March 2011 led to extensive fuel damage, including possible fuel melting, slumping, and relocation at the affected reactors. A so-called feed-and-bleed mode of reactor cooling was initially established to remove decay heat. The plan was to eventually switch over to a recirculation cooling system. Failure of feed and bleed was a possibility during the interim period. Furthermore, even if recirculation was established, there was a possibility of its subsequent failure. Decay heat has to be sufficiently removed to prevent further core degradation. To understand the possible evolution of the accident conditions and to have a tool for potential future hypothetical evaluations of accidents at other nuclear facilities, a Markov model of the state of the reactors was constructed in the immediate aftermath of the accident and was executed under different assumptions of potential future challenges. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accident. The work began in mid-March and continued until mid-May 2011. The analysis had the following goals: (1) To provide an overall framework for describing possible future states of the damaged reactors; (2) To permit an impact analysis of 'what-if' scenarios that could lead to more severe outcomes; (3) To determine approximate probabilities of alternative end-states under various assumptions about failure and repair times of cooling systems; (4) To infer the reliability requirements of closed loop cooling systems needed to achieve stable core end-states and (5) To establish the importance for the results of the various cooling system and physical phenomenological parameters via sensitivity calculations.

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-15

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

  15. Fission product chemistry in severe nuclear reactor accidents

    International Nuclear Information System (INIS)

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

  16. Accomplishments and challenges of the severe accident research

    International Nuclear Information System (INIS)

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

  17. PLINIUS-2. A new versatile platform for severe accident assessments

    International Nuclear Information System (INIS)

    The current PLINIUS platform at CEA Cadarache is a coherent set of experimental facilities devoted to knowledge improvements on corium behavior in severe accident conditions, model developments and code validation in view of performing reactor case calculations. In the frame of recent National and International Programs it provided numerous experimental results used for generation 2 and 3 reactors concerning major phenomena occurring during severe accident like Fuel Coolant Interaction (FCI) and Molten Corium Concrete Interaction (MCCI) as well as new data for thermophysical and thermochemical properties of the molten corium. One specific characteristic of this experimental platform is to produce significant masses of prototypical material (containing depleted uranium) molten at temperatures above 3000K and to carry out experiments with pools containing both metals and oxides, simulating most of the phenomena occurring in case of severe accident, except the non-thermal effects of radioactive decay. Moreover, specific measurements techniques have been developed giving a high level of accuracy of the results and allowing the assessment of analytical models like the fragmentation of the corium in the water. The mass of usable corium is limited to about 50 kg with the current furnaces and, even though some corium interaction experiments are already carried out in PLINIUS platform for generation 4 sodium fast reactors, sodium cannot be used in the current facility. Therefore, CEA has decided the building of a new large scale prototypic corium facility to extend its current PLINIUS corium platform. This new versatile facility, called PLINIUS 2, shall be devoted to both water and sodium-cooled reactor severe accident experimental simulation. This corium platform will cover all the R and D needs of the existing reactors, of new light water reactor designs and of the ASTRID project. It will be composed of a furnace able to melt 200 to 500 kg of corium including nuclear fuel

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-02-01

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

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

    International Nuclear Information System (INIS)

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

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

    Energy Technology Data Exchange (ETDEWEB)

    Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

    2015-09-01

    The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

  2. Severe head injury caused by motorcycle traffic accident

    Institute of Scientific and Technical Information of China (English)

    李钢

    1999-01-01

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

  3. A step forward towards understanding the severe accident evolution in Fukushima: Parametric analyses of Unit-1 with MELCOR 1.8.6

    International Nuclear Information System (INIS)

    This paper synthesizes the major progress achieved by CIEMAT in this international environment. The specific target pursued is the comprehension of the severe accident evolution. particularly the progression followed in the unit 1 of Fukushima Daiichi. To do so, the MELCOR 1.8.6 code has been used to model a Fukushima-like SBO sequence in a BWR-Mark 1 reactor. Particular emphasis has been given to understand the potential impact of systems like the isolation condenser (IC) and the over-pressure protection system.

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

    International Nuclear Information System (INIS)

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

  5. Numerical Study of Severe Accidents on Containment Venting Conditions

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

  6. Numerical Study of Severe Accidents on Containment Venting Conditions

    International Nuclear Information System (INIS)

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

  7. Advances in operational safety and severe accident research

    Energy Technology Data Exchange (ETDEWEB)

    Simola, K. [VTT Automation (Finland)

    2002-02-01

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

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

    International Nuclear Information System (INIS)

    In case of a severe accident in a nuclear reactor, very complex physical and chemical phenomena would occur. Parallel to the development of mechanistic and scenario codes, experiments are needed to determine key phenomena and coupling, develop and qualify specific models, validate codes. Experiments with prototypic corium are performed to check the results obtained with corium-simulant materials and identify possible differences. In this context CEA has undertaken a large program on severe accidents with prototypic corium. In this paper, we discuss some specificities of the prototypic corium: 1) Spreading: experiments with simulant mixtures and prototypic corium performed in the VULCANO facility showed a behaviour involving gas formation during melt spreading. 2) Corium pool: the presence of miscibility gap in the U-Zr-O ternary system for liquid phases and the high density of uranium oxides affect solidification paths, stratification and/or macro-segregation. 3) Corium concrete interaction: the possible reactions between uranium oxide and concrete oxides are specific in terms of thermodynamics and kinetics. For instance, the limited solubility of uranium in zircon can lead to the formation of the solid solution called ''chernobylite'' (Ux,Zr1-x)SiO4 which is important for the long term behaviour (fission product release, handling,..) of solidified corium. 4) Fuel Coolant Interaction: experiments in the KROTOS facility have shown important differences of behaviour between molten alumina and molten 80%wt UO2 + 20%wt ZrO2, the latter inducing less violent explosions than the former

  9. Advances in operational safety and severe accident research

    International Nuclear Information System (INIS)

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

  10. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. 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)

    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

  14. Shipping container response to three severe railway accident scenarios

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. 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.

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

    International Nuclear Information System (INIS)

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

  18. BNL severe accident sequence experiments and analysis program

    International Nuclear Information System (INIS)

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

  19. Mitigation of Severe Accident Consequences Using Inherent Safety Principles

    International Nuclear Information System (INIS)

    The safety challenges associated with sodium-cooled fast reactors have been recognized since the beginning of nuclear power and include the high power density in the core, the need for a reactor coolant and heat transfer system with high heat removal capability, the variation of power across the core requiring the use of ducted assemblies, and the condition that the fuel is not in the most neutronically reactive configuration during normal operation such that relocation can result in positive reactivity excursions, even possibly exceeding prompt critical conditions and energetic events. The potential for accidents with such severe consequences has been a negative factor with respect to the use of the sodium-cooled fast reactor. With the development of inherent safety principles, including favorable reactivity feedback, natural circulation cooling, and design choices resulting in favorable dispersive characteristics for failed fuel, it is possible to greatly increase the level of safety, to the point where it is highly unlikely, or perhaps even not possible, for accidents to result in releases of hghly radioactive materials to the containment or the surrounding environment. (author)

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

    OpenAIRE

    Högberg, Lars

    2013-01-01

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

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

    International Nuclear Information System (INIS)

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

  2. Implementation of Severe Accident Management Strategy at the Loviisa NPP

    International Nuclear Information System (INIS)

    A comprehensive severe accident management (SAM) strategy has been developed by Fortum for the Loviisa NPP in Finland. The strategy ensures reliable prevention and mitigation of containment - threatening phenomena, and it is built around a set of SAM safety functions. This paper focusses on the implementation status of the new SAM approach. We describe how and to what extent the modifications with regards to containment isolation, primary system depressurization, hydrogen mitigation, in-vessel retention of corium, and long-term containment cooling have been carried out. When implementing SAM, it was also necessary to modify the emergency response organisation to include a SAM support team. SAM guidelines, procedures and a SAM Handbook have been written. The automatic containment isolation function has been studied carefully within the SAM project. A successful isolation function is of paramount importance, when radioactive releases from the core can be expected to occur soon. Certain modifications have been carried out so that it is now possible to manually actuate missing isolation signals and to lock isolation status. New local control centres have been built to enable manual closure of certain isolation valves. Several new containment leak-tightness measurements have been installed. New depressurization valves, manually operated relief valves, were installed in 1996 for primary system depressurization purposes. The modifications to the ice condenser doors have been carried out in the years 2000 and 2001. Passive auto-catalytic recombiners have been successfully field-tested in the Loviisa containment atmosphere. We aim for installation in the year 2002. The locations of the glow plugs are being updated in a currently ongoing project. In-vessel retention of molten corium through external cooling of the reactor pressure vessel required certain plant modifications e.g. in order to guarantee access of water to the RPV wall. Most significantly, the support structures

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

    International Nuclear Information System (INIS)

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

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

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2012-01-01

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

  5. Accident analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant with the SAMPSON severe accident code. (1) Improvement of debris relocation model

    International Nuclear Information System (INIS)

    SAMPSON was designed as a large scale simulation system of inter-connected hierarchical modules covering a wide spectrum of scenarios ranging from normal operation to severe accidents. The code was validated by a wide range of analyses for separate-effect tests, and integral tests mainly through participation in the Organisation for Economic Co-operation and Development projects. In the previous analysis of TEPCO’s Fukushima Daiichi Nuclear Power Plant (1F) with the SAMPSON code, melt retention at a core plate was assumed based on observations after the Three Mile Island Unit 2 accident. The melt relocation to the core plate occurred when the water level was below the core plate in the SAMPSON analysis of the 1F accident. Therefore debris relocation phenomena were investigated using the Molten Core Relocation Analysis (MCRA) module of SAMPSON. The detailed model of the MCRA module was applied to the XR2-1 BWR metallic relocation experiment first. Molten material in the control rod area accumulated on the velocity limiter in the XR2-1 experiment and this phenomenon was reproduced by the SAMPSON analysis. A part of the molten metal fell directly through the inlet orifice in both the XR2-1 experiment and the SAMPSON analysis. Then the detailed model of the MCRA module was applied to the relocation phenomena of actual fuel bundles. The molten material accumulation on the velocity limiter and direct falling of the molten material through the inlet orifice were also observed in the analysis of actual fuel bundles. Based on the observations described above, MCRA noding for the system calculation was modified as follows. (1) The velocity limiters and control guide tubes were newly taken into account. (2) The flow path of debris was modified so that the molten materials could go to the lower plenum after passing through the inlet orifice without forced accumulation at the core plate. (author)

  6. Influence diagrams and decision trees for severe accident management

    International Nuclear Information System (INIS)

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

  7. Regulatory analyses for severe accident issues: an example

    International Nuclear Information System (INIS)

    A study has been performed as part of a program to establish methods for incorporation of information from a broad range of research programs, particularly those which generate probabilistic risk information, and to develop suitable presentation formats for providing guidance to decisionmakers on issues related to severe accidents. The study addresses issues related to information availability, content, and presentation formats for use in the regulatory decisionmaking process. The approach employed to address these issues was to perform an example regulatory analysis on representative topics of interest using available technical information. The issue examined in the example analysis is the implementation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The example demonstrates many of the problems which will be encountered as probabilistic information from ongoing programs is incorporated into the regulatory decisionmaking process

  8. Introduction of severe accidents mitigating systems used in TNPS

    International Nuclear Information System (INIS)

    Tianwan Nuclear Power Plant (TNPS) is a Russian WWER nuclear power plant (NPP). It is an improved type of the WWER 1000/V320 series. Also this NPP is the most advanced of the operated NPP in recent China. Many advanced safety concepts are used in the design of this NPP, so the safety level of TNPS is better than most of the NPPs in the world. Some systems designed in TNPS to mitigate the severe accidents impacts are firstly used in China, even in the world. These systems include Core Catcher, Reactor inspection shaft water emergency use system, Emergency gas removal system, Containment hydrogen removal system, and so on. This paper focus on the introduction of components and working principles of these systems. (author)

  9. A synthesis of hydrogen behaviour in severe reactor accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2001-07-01

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

  13. Experimental results of the CORA test program on the LWR fuel element behavior in severe reactor accidents

    International Nuclear Information System (INIS)

    In the framework of the CORA program the chemical interactions among fuel element (core) materials that may occur with increasing temperature up to complete melting have been examined. The high-temperature material behavior of PWR, BWR, and VVER-1000 fuel rod bundles has been studied in large-scale integral experiments and extensive separate-effects tests. In many cases, the reaction products are liquid at temperatures above 1200 C or have lower eutectic melting points than their original components. This results in a relocation of liquefied components, often far below their original melting points. Control rod materials can separate from fuel materials by a non-coherent stage-by-stage relocation process; this may cause recriticality problems during flooding of a partially degraded core with unborated water. Similarly, molten unoxidized Zircaloy cladding can relocate away from the decladded UO2 fuel rods. Significant relocation of UO2 dissolved in molten unoxidized Zircaloy can begin at the Zircaloy melting temperature (1760 C), about 1000 K below the melting point of UO2. Quenching (flooding) of the degraded bundles results in locally enhanced Zircaloy/steam reactions causing a renewed temperature rise, a meltdown of materials, and an additional strong H2 generation. The experimental results have contributed substantially to the understanding of the high-temperature core material behavior in severe reactor accidents, and provided a unique data base for the development, improvement, and validation of material-behavior models and severe accident system codes. (orig.)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  16. Severe Accident Sequence Analysis Program: Anticipated transient without scram simulations for Browns Ferry Nuclear Plant Unit 1

    International Nuclear Information System (INIS)

    An analysis of five anticipated transients without scram (ATWS) was conducted at the Idaho National Engineering Laboratory (INEL). The five detailed deterministic simulations of postulated ATWS sequences were initiated from a main steamline isolation valve (MSIV) closure. The subject of the analysis was the Browns Ferry Nuclear Plant Unit 1, a boiling water reactor (BWR) of the BWR/4 product line with a Mark I containment. The simulations yielded insights to the possible consequences resulting from a MSIV closure ATWS. An evaluation of the effects of plant safety systems and operator actions on accident progression and mitigation is presented

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  2. EPRTM engineered features for core melt mitigation in severe accidents

    International Nuclear Information System (INIS)

    For the prevention of accident conditions, the EPRTM relies on the proven 3-level safety concepts inherited from its predecessors, the French 'N4' and the German 'Konvoi' NPP. In addition, a new, fourth 'beyond safety' level is implemented for the mitigation of postulated severe accidents (SA) with core melting. It is aimed at preserving the integrity of the containment barrier and at significantly reducing the frequency and magnitude of activity releases into the environment under such extreme conditions. Loss of containment integrity is prevented by dedicated design measures that address short- and long-term challenges, like: the melt-through of the reactor pressure vessel under high internal pressure, energetic hydrogen/steam explosions, containment overpressure failure, and basemat melt-through. The EPRTM SA systems and components that address these issues are: - the dedicated SA valves for the depressurization the primary circuit, - the provisions for H2 recombination, atmospheric mixing, steam dilution, - the core melt stabilization system, - the dedicated SA containment heat removal system. The core melt stabilization system (CMSS) of the EPRTM is based on a two-stage ex-vessel approach. After its release from the RPV the core debris is first accumulated and conditioned in the (dry) reactor pit by the addition of sacrificial concrete. Then the created molten pool is spread into a lateral core catcher to establish favorable conditions for the later flooding, quenching and cooling with water passively drained from the Internal Refueling Water Storage Tank. Long-term heat removal from the containment is achieved by sprays that are supplied with water by the containment heat removal system. Complementing earlier publications focused on the principle function, basic design, and validation background of the EPRTM CMSS, this paper describes the state achieved after detailed design, as well as the technical solutions chosen for its main components, including their

  3. 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.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  5. Coupled thermal-hydraulic/aerosol transport analysis capability for severe accidents

    International Nuclear Information System (INIS)

    Fission product transport and thermal-hydraulic phenomena, occurring during the in-vessel phase of postulated accident progression, affect each other directly and significantly. It is important to couple the calculation of these processes to obtain accurate estimates of the magnitude and the timing of the release, of initial and revolatilized fission products from the primary system of a PWR, or the vessel of a BWR. The Electric Power Research Institute research efforts to obtain a functional and sophisticated coupled thermal-hydraulic/aerosol transport analysis capability are described. These are based on coupling the codes CORMLT, PSAAC (Primary System Accident Analysis Code) and RAFT (Reactor Aerosol Formation and Transport), which have been under development since 1982. Summary descriptions of these codes are also provided. (author)

  6. Scaling material effects in late phases of severe accidents

    International Nuclear Information System (INIS)

    Several complex physical phenomena are involved in the late phases of severe accidents of LWRs. The weak understanding of aspects related to the corium behaviour in the melting interval leads to difficulties in the definition and extrapolation of experiments that are performed with simulant materials. Therefore real material experiments at realistic scale are preferred. However, real material experiments are very expensive and are limited in size. Recent developments at CEA addressed the problem of the material behaviour in the melting interval. Models have been developed and qualified on the basis of the analysis of available experimental results. These models are based on a coupling between physico-chemistry and thermal hydraulics for the description of corium pools, long term corium-concrete or corium-ceramic interactions, corium spreading, low-volatile fission products release, and corium properties (viscosity). On the basis of the phenomenological understanding and of the model approach developed, it is possible to derive scaling criteria for material effects in support of the definition of simulant material experiments to address the open issues that will be mentioned. (author)

  7. A study on the establishment of severe accident experimental facility

    International Nuclear Information System (INIS)

    Significant progress has been achieved during this year of the project. Planned DCH experiments on the sensitivity of the cavity geometry factors and the cavity capture volume effects were performed using the HPME facility for Kori-1/2 and YGN-3/4 cavity scale models. The Crust Formation Test Facility has been completed. Preliminary calculations were performed to predict test results. The experiments of the crust formation on the simulant and its heat transfer characteristic were performed to investigate the effects of coolant injection methods, bottom heating boundary surface temperatures, coolant temperatures and coolant flow rates. The design of the FCI Test Facility has been completed and the procurement of the materials is in progress. Also, the steam condensation experiment on the vertical containment walls and the research on the development of measuring techniques of the particle sizes and velocities are in progress as planned. Through international research collaboration with USNRC and CEA Cadarache, information of the experimental research on the severe fuel damage has been gathered and analyzed. Preliminary planning of the second phase tests has been launched this year. This study proposes the scope of the second phase and the strategy to implement the proposed second phase experimental program. This study also proposes a strategy to establish building blocks and infrastructure for the severe accident research in Korea. (Author)

  8. Scaling material effects in late phases of severe accidents

    Energy Technology Data Exchange (ETDEWEB)

    Seiler, J.M.; Froment, K. [CEA Grenoble, Dept. de Thermohydraulique et de Physique, 38 (France)

    2001-07-01

    Several complex physical phenomena are involved in the late phases of severe accidents of LWRs. The weak understanding of aspects related to the corium behaviour in the melting interval leads to difficulties in the definition and extrapolation of experiments that are performed with simulant materials. Therefore real material experiments at realistic scale are preferred. However, real material experiments are very expensive and are limited in size. Recent developments at CEA addressed the problem of the material behaviour in the melting interval. Models have been developed and qualified on the basis of the analysis of available experimental results. These models are based on a coupling between physico-chemistry and thermal hydraulics for the description of corium pools, long term corium-concrete or corium-ceramic interactions, corium spreading, low-volatile fission products release, and corium properties (viscosity). On the basis of the phenomenological understanding and of the model approach developed, it is possible to derive scaling criteria for material effects in support of the definition of simulant material experiments to address the open issues that will be mentioned. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1998-03-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1990-12-01

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

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

    Institute of Scientific and Technical Information of China (English)

    2007-01-01

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

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

    International Nuclear Information System (INIS)

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

  14. 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.

  15. BNL severe accident sequence experiments and analysis program

    International Nuclear Information System (INIS)

    A major source of containment pressurization during severe accidents is the transfer of stored energy from the hot core material to available cooling water. One mode of thermal interaction involves the quench of superheated beds of debris which could be present in the reactor cavity following melt-through or failure of the reactor vessel. This work supports development of models of superheated bed quench phenomena which are to be incorporated into containment analysis computer codes such as MARCH, CONTAIN, and MEDICI. A program directed towards characterization of the behavior of superheated debris beds has been completed. This work addressed the quench of superheated debris which is postulated to exist in the reactor cavity of a PWR following melt ejection from the primary system. The debris is assumed to be cooled by a pool of water overlying the bed of hot debris. This work has led to the development of models to predict rate of steam generation during the quench process and, in addition, the ability to assess the coolability of the debris during the transient quench process. A final report on this work has been completed. This report presents a brief description of some relevant results and conclusions. 15 refs

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

    International Nuclear Information System (INIS)

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

  17. SWR-1000 concept on control of severe accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2002-01-01

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

  20. Mechanical behaviour of reactor pressure vessel in severe accident

    International Nuclear Information System (INIS)

    The article describes the main achievements in developing methodology for analysing mechanical behaviour of pressure vessels with and without penetrations at high temperatures related to accident scenarios. Validation and applications of the methodology are presented. (orig.)

  1. Development of the MIDAS GUI environment for severe accident management and analyses

    International Nuclear Information System (INIS)

    MIDAS is being developed at KAERI as an integrated severe accident analysis code with existing model modification and new model addition. Also restructuring of the data transfer scheme is going on to improve user's convenience. In this paper, various MIDAS GUI systems which are input management system IEDIT, variable plotting system IPLOT, severe accident training simulator SATS, and online guidance module HyperKAMG, are introduced. In addition, detail functions and usage of these systems for severe accident management and analyses are described

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

    OpenAIRE

    Min Yoo; Sung Min Shin; Hyun Gook Kang

    2014-01-01

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

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

    Institute of Scientific and Technical Information of China (English)

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

    2012-01-01

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

  4. Severe accident progression perspectives based on IPE results

    Energy Technology Data Exchange (ETDEWEB)

    Lehner, J.R.; Lin, C.C.; Pratt, W.T. [Brookhaven National Laboratory, Upton, NY (United States); Drouin, M. [Nuclear Regulatory Commission, North Bethesda, MD (United States)

    1996-08-01

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

  5. Developing and validating severe accident management guidelines using SAMPSON-RELAP/SCDAPSIM.MOD3.4

    International Nuclear Information System (INIS)

    The development and validation of Severe Accident Management Guidelines (SAMGs) must consider complex thermal-hydraulic and severe accident phenomena. Yet, many of the simplified integral Severe Accident codes, that have been used widely to develop SAMGs in Europe, Asia, and the United States, cannot accurately predict many of these complex interactions. By contrast, detailed codes such as SAMPSON-RELAP/SCDAPSIM have shown, through comparison with the TMI-2 accident and experiments, that they can predict such complex behavior. This paper describes the merger of SAMPSON with RELAP/SCDAPSIM/MOD3.4, reviews the severe accident phenomena important for Severe Accident Management, and then describes the potential impact of using SAMPSON-RELAP/SCDAPSIM on the development and validation of SAMGs. A companion paper, being presented at this conference provides an example of the application of SAMPSON-RELAP/SCDAPSIM for the development and validation of a SAMG for a Nuclear Power Plant. (authors)

  6. 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.

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

    International Nuclear Information System (INIS)

    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

  8. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

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

  9. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-11-01

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

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

    International Nuclear Information System (INIS)

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

  15. Analysis on the severe accidents in KSTAR tokamak

    Energy Technology Data Exchange (ETDEWEB)

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

    2003-11-15

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

  16. Severe accident mitigation features of the economic simplified boiling water reactor

    International Nuclear Information System (INIS)

    This paper provides an overview of the Economic Simplified Boiling Water Reactor (ESBWR)severe accident mitigation systems. The major severe accident types are described and the systems credited for mitigating the severe accidents are discussed, including the Basemat Internal Melt Arrest Coolability (BiMAC) device, the Passive Containment Cooling System (PCCS), and the advantages of suppression pool water for scrubbing during containment venting. The ruggedness of the containment and reactor building designs for accommodating beyond design accident conditions is also discussed. (author)

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

    OpenAIRE

    Jan Christian Kaiser

    2012-01-01

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

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

    OpenAIRE

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

    2014-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1999-07-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Sugimoto, Jun [ed.

    1998-05-01

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

  1. Management of a severe accident on a pressurised water reactor in France

    International Nuclear Information System (INIS)

    This brief document defines what a severe accident is on a nuclear reactor, indicates the different failure modes which have been defined (vapour explosion in the reactor vessel, hydrogen explosion, and so on). It describes the management of a core fusion accident for pressurized water reactors, for which a guide has been designed, the GIAG (intervention guide for a severe accident situation). The principles of such an intervention are described, and then the approach for an EPR reactor

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

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

    International Nuclear Information System (INIS)

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

  5. Use of RETAIN-S for calculation of fission product transport: Comparison with experimental results and applications to BWR meltdown accidents

    International Nuclear Information System (INIS)

    RETAIN-S is a code for calculating fission product transport in a multicompartment system. The fission products can occur in the form of aerosols and vapours, and revaporization due to surface heating is taken into consideration. The aerosol model uses log-normal approximation for size distribution. The differences between RETAIN-S and the Industry Degraded Core Rulemaking (IDCOR) version of RETAIN are mentioned briefly. The method for calculating condensation of vapours and vaporization due to heating is described. The results of test calculations made on the Surry AB sequence defined by the Group of Experts on the Source Term (GREST) are given and compared with corresponding NAUA results. It is concluded that the degree of agreement is strongly dependent on the model used for gravitational agglomeration efficiency. The calculations made on a Marviken-V experiment are also reported. It is shown that the results are dependent on the possible condensation of fission product vapour on the walls and on aerosols. In addition, the results are given for a BWR meltdown sequence characterized by a large loss-of-coolant accident with loss of all AC power. One important conclusion is that the source term after containment break is mainly determined by the possible re-evaporation of fission products. (author)

  6. MELCOR/SNAP analysis of Chinshan (BWR/4) Nuclear Power Plant spent fuel pool for the similar Fukushima accident

    International Nuclear Information System (INIS)

    Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP event occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool, by using MELCOR 2.1 and SNAP 2.2.7 codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP spent fuel pool (SFP). There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP MELCOR/SNAP model. And the transient analysis under the SFP cooling system failure condition was performed. Besides, in order to study the detailed thermal-hydraulic performance of this transient, TRACE was used in this analysis. CFD data from INER report was used to compare with the results of MELCOR and TRACE. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Besides, the animation model of Chinshan NPP SFP was presented using the animation function of SNAP with MELCOR analysis results. (author)

  7. BWR 90 - the advanced BWR of the 1990s

    International Nuclear Information System (INIS)

    The future global role of nuclear power will be determined by its ability to provide economical and safe energy. Nuclear power, like any other substantial contributor to the world's energy needs, must be generated at an acceptable cost and with negligible environmental effects. Besides, it must achieve and maintain a socially reasonable level of public acceptance, which in turn is not necessarily governed by rational assessments of the true safety and environmental impact of nuclear power. The ABB Atom approach to this situation can best be characterized as a 'cautious evolution'; for the next decade the company will largely base its offerings to the market on its 'evolutionary' light water reactor design, the BWR 90. This design builds closely on the experience from successful construction and operation of its predecessor, the BWR 75 design. In 1995 and 1996, plants of this design achieved an average load factor greater than the 87% set by EUR; the two BWR units at Olkiluoto in Finland are among the very best performing plants in the world, with an average load factor of 94% over the last 7 years. The continued LWR design development focuses on meeting requirements from utilities as well as new regulatory requirements. A particular emphasis is put on the consequences of severe accidents; there shall be no large releases to the environment. Other design improvements involve: all-digital I and C systems and enhanced human factors engineering to improve work environment for operators, optimization of buildings and containment design to decrease construction time and costs, and selection of materials as well as maintenance and operating procedures to even further reduce occupational radiation exposures. (orig.)

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

    International Nuclear Information System (INIS)

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

  9. 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

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

    International Nuclear Information System (INIS)

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

  11. Evaluation of severe accident risks: Quantification of major input parameters

    International Nuclear Information System (INIS)

    This report records part of the vast amount of information received during the expert judgment elicitation process that took place in support of the NUREG-1150 effort sponsored by the U.S. Nuclear Regulatory Commission. The results of the Containment Loads and Molten Core/Containment Interaction Expert Panel Elicitation are presented in this part of Volume 2 of NUREG/CR-4551. The Containment Loads Expert Panel considered seven issues: (1) hydrogen phenomena at Grand Gulf; (2) hydrogen burn at vessel breach at Sequoyah; (3) BWR reactor building failure due to hydrogen; (4) Grand Gulf containment loads at vessel breach; (5) pressure increment in the Sequoyah containment at vessel breach; (6) loads at vessel breach: Surry; and (7) pressure increment in the Zion containment at vessel breach. The report begins with a brief discussion of the methods used to elicit the information from the experts. The information for each issue is then presented in five sections: (1) a brief definition of the issue, (2) a brief summary of the technical rationale supporting the distributions developed by each of the experts, (3) a brief description of the operations that the project staff performed on the raw elicitation results in order to aggregate the distributions, (4) the aggregated distributions, and (5) the individual expert elicitation summaries. The Molten Core/Containment Interaction Panel considered three issues. The results of the following two of these issues are presented in this document: (1) Peach Bottom drywell shell meltthrough; and (2) Grand Gulf pedestal erosion. 89 figs., 154 tabs

  12. Reactor safety study. An assessment of accident risks in U. S. commercial nuclear power plants. Appendix VI. Calculation of reactor accident consequences. [PWR and BWR

    Energy Technology Data Exchange (ETDEWEB)

    1975-10-01

    Information is presented concerning the radioactive releases from the containment following accidents; radioactive inventory of the reactor core; atmospheric dispersion; reactor sites and meteorological data; radioactive decay and deposition from plumes; finite distance of plume travel; dosimetric models; health effects; demographic data; mitigation of radiation exposure; economic model; and calculated results with consequence model.

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

    International Nuclear Information System (INIS)

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

  14. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    Energy Technology Data Exchange (ETDEWEB)

    Carbajo, J.J. [Oak Ridge National Lab., TN (United States)

    1993-09-01

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences.

  15. Severe accident source term characteristics for selected Peach Bottom sequences predicted by the MELCOR Code

    International Nuclear Information System (INIS)

    The purpose of this report is to compare in-containment source terms developed for NUREG-1159, which used the Source Term Code Package (STCP), with those generated by MELCOR to identify significant differences. For this comparison, two short-term depressurized station blackout sequences (with a dry cavity and with a flooded cavity) and a Loss-of-Coolant Accident (LOCA) concurrent with complete loss of the Emergency Core Cooling System (ECCS) were analyzed for the Peach Bottom Atomic Power Station (a BWR-4 with a Mark I containment). The results indicate that for the sequences analyzed, the two codes predict similar total in-containment release fractions for each of the element groups. However, the MELCOR/CORBH Package predicts significantly longer times for vessel failure and reduced energy of the released material for the station blackout sequences (when compared to the STCP results). MELCOR also calculated smaller releases into the environment than STCP for the station blackout sequences

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. The scrubbing of fission product aerosols in LWR water pools under severe accident conditions

    International Nuclear Information System (INIS)

    The analysis of source teams for core damage accidents requires that the accident progression, the plant thermal hydraulic response and the associated radionuclide release and transport be analyzed for the identified escape pathways. These pathways are identified by examining the plant design, the events which constitute the hypothetical accident sequences, and the analyzed or assumed plant response. Many of the pathways currently identified for what are believed to be the risk significant sequences involve pathway segments through water pools. Specifically, transport through the BWR pressure suppression pool and the PWR quench tank represent examples of segments of the in-plant pathways for many risk significant sequences in these plants. In order to properly analyze the fission product transport for these sequences, it is therefore necessary to describe the effect of these water pools on the steam/non-condensible gas mixtures and associated entrained radionuclides. This paper describes Phase I of the experimental pool scrubbing work that is being conducted in conjunction with the development and validation of the SUPRA code

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  2. Thermal Hydraulic design parameters study for severe accidents using neural networks

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-12-31

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

  3. Development and application of calculational theoretical methods for analysis of the RBMK reactor severe accidents

    International Nuclear Information System (INIS)

    One studied high-improbable reactor emergencies that may result in a high consequence accident. To control these accidents and to mitigate their consequences one should study and analyze similar emergencies via detailed computer simulation. Application of foreign and Russian codes for RBMK type reactor should be associated with their supplementary verification. In that context one elaborated the table list of processes for supplementary verification of thermohydraulic models of codes designed to analyze severe accidents

  4. Study of the Severity of Accidents in Tehran Using Statistical Modeling and Data Mining Techniques

    Directory of Open Access Journals (Sweden)

    Hesamaldin Razi

    2013-01-01

    Full Text Available AbstractBackgrounds and Aims: The Tehran province was subject to the second highest incidence of fatalities due to traffic accidents in 1390. Most studies in this field examine rural traffic accidents, but this study is based on the use of logit models and artificial neural networks to evaluate the factors that affect the severity of accidents within the city of Tehran.Materials and Methods: Among the various types of crashes, head-on collisions are specified as the most serious type, which is investigated in this study with the use of Tehran’s accident data. In the modeling process, the severity of the accident is the dependent variable and defined as a binary covariate, which are non-injury accidents and injury accidents. The independent variables are parameters such as the characteristics of the driver, time of the accident, traffic and environmental characteristics. In addition to the prediction accuracy comparison of the two models, the elasticity of the logit model is compared with a sensitivity analysis of the neural network.Results: The results show that the proposed model provides a good estimate of an accident's severity. The explanatory variables that have been determined to be significant in the final models are the driver’s gender, age and education, along with negligence of the traffic rules, inappropriate acceleration, deviation to the left, type of vehicle, pavement conditions, time of the crash and street width.Conclusion: An artificial neural network model can be useful as a statistical model in the analysis of factors that affect the severity of accidents. According to the results, human errors and illiteracy of drivers increase the severity of crashes, and therefore, educating drivers is the main strategy that will reduce accident severity in Iran. Special attention should be given to a driver’s age group, with particular care taken when they are very young.

  5. Study on severe accidents and countermeasures for WWER-1000 reactors using the integral code ASTEC

    International Nuclear Information System (INIS)

    The research field focussing on the investigations and the analyses of severe accidents is an important part of the nuclear safety. To maintain the safety barriers as long as possible and to retain the radioactivity within the airtight premises or the containment, to avoid or mitigate the consequences of such events and to assess the risk, thorough studies are needed. On the one side, it is the aim of the severe accident research to understand the complex phenomena during the in- and ex-vessel phase, involving reactor-physics, thermal-hydraulics, physicochemical and mechanical processes. On the other side the investigations strive for effective severe accident management measures. This paper is focused on the possibilities for accident management measures in case of severe accidents. The reactor pressure vessel is the last barrier to keep the molten materials inside the reactor, and thus to prevent higher loads to the containment. To assess the behaviour of a nuclear power plant during transient or accident conditions, computer codes are widely used, which have to be validated against experiments or benchmarked against other codes. The analyses performed with the integral code ASTEC cover two accident sequences which could lead to a severe accident: a small break loss of coolant accident and a station blackout. The results have shown that in case of unavailability of major active safety systems the reactor pressure vessel would ultimately fail. The discussed issues concern the main phenomena during the early and late in-vessel phase of the accident, the time to core heat-up, the hydrogen production, the mass of corium in the reactor pressure vessel lower plenum and the failure of the reactor pressure vessel. Additionally, possible operator's actions and countermeasures in the preventive or mitigative domain are addressed. The presented investigations contribute to the validation of the European integral severe accidents code ASTEC for WWER-1000 type of reactors

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  9. Relevant scenarios and uncertainty analysis of severe accidents in the U.S. EPR

    International Nuclear Information System (INIS)

    As part of U.S. EPR design certification activities, AREVA has prepared analyses to support the US NRC's regulatory expectation with regard to the resolution of several severe accident safety issues identified in SECY 93-087. To address the large uncertainties associated with severe accident progression, AREVA NP has developed and applied a best-estimate plus uncertainty methodology to the analysis of severe accidents. The uncertainty methodology considers a broad spectrum of phenomenological and process uncertainties. Unique among the uncertainty parameters considered is the sampling of event sequence (i.e., scenario type). (authors)

  10. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil is under development. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of a severe accidents. (author)

  11. Development of internal CRD for next generation BWR

    International Nuclear Information System (INIS)

    In order to develop a competitive and high performance Next Generation BWR with fossil power plant, an internal CRD using a heatproof ceramics insulated coil has been developed. In case of a 1700MWe next generation BWR, the internal CRDs are installed in a RPV whose size is equivalent to the 1356 MWe ABWR, and there will be no space required for CRDs and CRD exchange under RPV. These advantages realize a compact PCV and reduced volume of a reactor building. Moreover, the internal CRDs eliminate penetration via a bottom flange of RPV, and lower installation level of RPV in a drywell. This brings further advantages of elimination of RIA (Reactivity Induced Accidents) caused by CR withdrawing under pressure boundary broken, and easy IVR (In Vessel Retention) by vessel bottom cooling in case of severe accidents. (authors)

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Kim, Dong Ha

    2006-08-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Hashimoto, Kazuichiro [ed.

    2000-11-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  19. A study on the use of neural network for severe accident management

    International Nuclear Information System (INIS)

    Based on the consensus that the course and consequence of a severe core damage accident can be greatly influenced by the operators' action, there have been extensive efforts to establish severe accident management program. A severe accident management process is essentially a sequence of decision making with a wide variety of available information under the highly uncertain condition, aimed at successful termination of accident progression or consequence minimization. For operators to take correct and timely accident management actions, they should be informed of the accident progression. Some key events, such as onset of core uncovery, core-melt initiation, reactor vessel lower head failure, containment failure, etc., act as landmarks for operators to make decisions in severe accident management process. Thus it is of critical importance to identify the timing at which such events occur in accident management. Unfortunately, it is difficult task partly due to phenomenological complexity and partly due to the lack of instrumentation reliability in severe accident environment, making the traditional procedural or rule-based approach inappropriate to be adopted to this end. Instead a technique, called artificial neural network, has been successfully applied to the similar problem domain out of various disciplines including nuclear industry. This paper presents a study on the application of a special kind of artificial neural network having the capability of recognizing time-varying patterns, called spatiotemporal network (STN), to the event timing prediction which is an important sub function of integrated computer supporting system for severe accident management. As the first trial, concentration was put on the identification of reactor vessel lower head failure which is considered the most critical events discriminating between so called in-vessel and ex-vessel accident management phases. Several sets of seven parameter signals from MAAP-based severe accident

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

    International Nuclear Information System (INIS)

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

  1. An analysis of LOCA sequences in the development of severe accident analysis DB

    International Nuclear Information System (INIS)

    Although a Level 2 PSA was performed for the Korean Standard Power Plants (KSNPs), and it considered the necessary sequences for an assessment of the containment integrity and source term analysis. In terms of an accident management, however, more cases causing severe core damage need to be analyzed and arranged systematically for an easy access to the results. At present, KAERI is calculating the severe accident sequences intensively for various initiating events and generating a database for the accident progression including thermal hydraulic and source term behaviours. The developed Database (DB) system includes a graphical display for a plant and equipment status, previous research results by knowledge-base technique, and the expected plant behaviour. The plant model used in this paper is oriented to the case of LOCAs related severe accident phenomena and thus can simulate the plant behaviours for a severe accident. Therefore the developed system may play a central role as an information source for decision-making for a severe accident management, and will be used as a training simulator for a severe accident management. (author)

  2. Development of the MIDAS GUI environment for severe accident management and analyses

    Energy Technology Data Exchange (ETDEWEB)

    Kim, K. R.; Park, S. H.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

    2004-07-01

    MIDAS is being developed at KAERI as an integrated severe accident analysis code with existing model modification and new model addition. Also restructuring of the data transfer scheme is going on to improve user's convenience. In this paper, various MIDAS GUI systems which are input management system IEDIT, variable plotting system IPLOT, severe accident training simulator SATS, and online guidance module HyperKAMG, are introduced. In addition, detail functions and usage of these systems for severe accident management and analyses are described.

  3. Heating of reactor pressure vessel bottom head and penetrations in a severe reactor accident; Reaktoripaineastian pohjan ja laepivientien kuumeneminen sydaemen sulamisonnettomuudessa

    Energy Technology Data Exchange (ETDEWEB)

    Ikonen, K. [VTT Energy, Espoo (Finland). Nuclear Energy

    1997-10-01

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

  4. Advanced BWR core component designs and the implications for SFD analysis

    International Nuclear Information System (INIS)

    Prior to the DF-4 boiling water reactor (BWR) severe fuel damage (SFD) experiment conducted at the Sandia National Laboratories in 1986, no experimental data base existed for guidance in modeling core component behavior under postulated severe accident conditions in commercial BWRs. This paper will present the lessons learned from the DF-4 experiment (and subsequent German CORA BWR SFD tests) and the impact on core models in the current generation of SFD codes. The DF-4 and CORA BWR test assemblies were modeled on the core component designs circa 1985; that is, the 8 x 8 fuel assembly with two water rods and a cruciform control blade constructed of B4C-filled tubelets. Within the past ten years, the state-of-the-art with respect to BWR core component development has out-distanced the current SFD experimental data base and SFD code capabilities. For example, modern BWR control blade design includes hafnium at the tips and top of each control blade wing for longer blade operating lifetimes; also water rods have been replaced by larger water channels for better neutronics economy; and fuel assemblies now contain partial-length fuel rods, again for better neutronics economy. This paper will also discuss the implications of these advanced fuel assembly and core component designs on severe accident progression and on the current SFD code capabilities

  5. BWR/5 Pressure-Suppression Pool Response during an SBO

    OpenAIRE

    Javier Ortiz-Villafuerte; Andrés Rodríguez-Hernández; Enrique Araiza-Martínez; Luis Fuentes-Márquez; Jorge Viais-Juárez

    2013-01-01

    RELAP/SCDAPSIM Mod 3.4 has been used to simulate a station blackout occurring at a BWR/5 power station. Further, a simplified model of a wet well and dry well has been added to the NSSS model to study the response of the primary containment during the evolution of this accident. The initial event leading to severe accident was considered to be a LOOP with simultaneous scram. The results show that RCIC alone can keep the core fully covered, but even in this case about 30% of the original liqui...

  6. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    Energy Technology Data Exchange (ETDEWEB)

    Hermsmeyer, S. [European Commission JRC, Petten (Netherlands). Inst. for Energy and Transport; Herranz, L.E.; Iglesias, R. [CIEMAT, Madrid (Spain); and others

    2015-07-15

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  7. Review of current severe accident management approaches in Europe and identification of related modelling requirements for the computer code ASTEC V2.1

    International Nuclear Information System (INIS)

    The severe accident at the Fukushima-Daiichi nuclear power plant (NPP) has led to a worldwide review of nuclear safety approaches and is bringing a refocussing of R and D in the field. To support these efforts several new Euratom FP7 projects have been launched. The CESAM project focuses on the improvement of the ASTEC computer code. ASTEC is jointly developed by IRSN and GRS and is considered as the European reference code for Severe Accident Analyses since it capitalizes knowledge from the extensive Euro-pean R and D in the field. The project aims at the code's enhancement and extension for use in Severe Accident Management (SAM) analysis of the NPPs of Generation II-III presently under operation or foreseen in the near future in Europe, spent fuel pools included. The work reported here is concerned with the importance, for the further development of the code, of SAM strategies to be simulated. To this end, SAM strategies applied in the EU have been compiled. This compilation is mainly based on the public information made available in the frame of the EU ''stress tests'' for NPPs and has been complemented by information pro-vided by the different CESAM partners. The context of SAM is explained and the strategies are presented. The modelling capabilities for the simulation of these strategies in the current production version 2.0 of ASTEC are discussed. Furthermore, the requirements for the next version of ASTEC V2.1 that is supported in the CESAM project are highlighted. They are a necessary complement to the list of code improvements that is drawn from consolidating new fields of application, like SFP and BWR model enhancements, and from new experimental results on severe accident phenomena.

  8. A review of iodine chemistry under severe accident conditions

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Codes for NPP severe accident simulation: development, validation and applications

    International Nuclear Information System (INIS)

    The software tools that describe various safety aspects of NPP with VVER reactor have been developed at the Nuclear Safety Institute of the Russian Academy of Sciences (IBRAE RAN). Functionally, the codes can be divided into two groups: the calculation codes that describe separate elements of NPP equipment and/or a group of processes and integrated software systems that allow solving the tasks of the NPP safety assessment in coupled formulation. In particular, IBRAE RAN in cooperation with the nuclear industry organizations has developed the integrated software package SOCRAT designed to analyze the behavior of NPP with VVER at various stages of beyond-design-basis accidents, including the stages of reactor core degradation and long-term melt retention in a core catcher. The general information about development, validation and applications of SOCRAT code is presented and discussed in the paper. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2004-08-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  17. Preliminary Study on Information Display System for Operator Support in Severe Accident

    International Nuclear Information System (INIS)

    This paper proposes an information display system dedicated to a severe accident and presents the conceptual requirements to support a prompt decision of operators under severe accident conditions. In this paper, SIDS, which mainly supports the decision-making of operator under severe accident conditions, was proposed and its conceptual requirements were presented to support a prompt decision of operators under severe accident conditions. Also the conceptual requirements regarding the functions, performance, and information display were established. Further study on designing additional systems and equipment for severe accidents will be accomplished in the future after analyzing the SAMG, severe scenarios, and equipment survivability from instrument sensor to I and C facilities. This paper discusses the issues related to the severe accident of Fukushima Daiichi Nuclear Power Electric Power, which occurred on March 11, 2011, due to the Pacific Ocean Earthquake and the ensuing tsunami generated by the Earthquake. As a result of this severe accident, a large amount of radioactive material was released. From an I and C(instrumentation and control) point of view, I and C equipment did not function sufficiently or was shut off due to the loss of power supplies. The operator could not react effectively under these conditions because it was difficult to obtain the appropriate information needed to take countermeasures. Most commercial nuclear power plants use an operator support system such as an SPDS(safety parameter display system) under normal and emergency conditions to support the prompt decision of the operator. However, SPDS is suitable for emergency conditions, not for severe accident conditions

  18. A training simulator to support the Loviisa VVER-440 severe accident management programme

    International Nuclear Information System (INIS)

    A simulation tool for training operators and technical support personnel for severe accidents is being developed at VTT. The system will be accomplished by implementing severe accident models into the APROS - Advanced Process Simulator - environment, which already includes a model of the Loviisa VVER-440 plant. The system development is closely coupled with the plant severe accident management programme. The Loviisa severe accident management programme consists of four high level actions: primary system depressurization, retention of molten core within the pressure vessel, hydrogen control and containment external spray cooling. The training system will at the first stage simulate in simple terms the key phenomena associated with these actions and their effect on the plant response. The paper describes the system objectives, outline and modelling philosophy

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

    Czech Academy of Sciences Publication Activity Database

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

    2014-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-02-15

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2013-07-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  7. 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.

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

    OpenAIRE

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

    2016-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  11. Development of a severe-accident simulator with a visual plant behavior display

    International Nuclear Information System (INIS)

    Severe-accident management is one of the important safety concerns of the nuclear industry and regulatory organizations. Mitsubishi Atomic Power and Mitsubishi Heavy Industries in Japan have developed a severe-accident simulator with the ability to display plant thermal-hydraulic behavior visually in order to develop operating guidelines and to use as an education and training tool. The main features of this simulator are described

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

    OpenAIRE

    MohammadJavad Jafari; Iraj Mohammadfam; Esmaeil Zarei

    2015-01-01

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  15. European expert network for the reduction of uncertainties in severe accident safety issues (EURSAFE)

    International Nuclear Information System (INIS)

    EURSAFE thematic network was a concerted action in the sixth framework programme of the European Commission. It established a large consensus among the main actors in nuclear safety on the severe accident issues where large uncertainties still subsist. The conclusions were derived from a first-of-kind phenomena identification and ranking tables (PIRT) on all aspects of severe accident also realised in the frame of the project. Starting from a list of all severe accident phenomena containing approximately 1000 entries and established by the twenty partner organisations, 106 phenomena were retained eventually as both important for safety and still lacking sufficient knowledge. Ultimately, 21 research areas for addressing these phenomena regrouped according to their similarities were identified. A networking structure for implementing and executing the necessary research was proposed, which promotes integration and harmonisation of the different national programmes. A severe accident database structure was proposed to ensure preservation of experimental data and enhanced communication for data exchange and use for severe accident codes assessment. The final product, named EURSAFE, is a website network, http://asa2.jrc.it/eursafe, connecting nodes located at partner sites. As the result of an action involving R and D governmental institutions, regulatory bodies, nuclear industry, utilities and universities from six EU Member States (Finland, France, Germany, Spain, Sweden, UK) plus JRC, three European third countries (Czech Republic, Hungary, Switzerland), and USA, EURSAFE represents a significant step towards harmonisation and credibility of the approaches, and resolution of the remaining severe accident issues

  16. Boiling heat transfer phenomenon base on the event of loca and severe accident

    International Nuclear Information System (INIS)

    Research and development base on TMI-2 NPP accident mostly directed to vessel and core performance. The majority of research was conducted which aimed on boiling heat transfer phenomenon, begin by loss of coolant accident (LOCA) until severe accident, in which core meltdown. Study on boiling heat transfer has been done by simulation on core bottom re-flooding process and a narrow gap cooling. The results of experimental research which was conducted by BATAN concerning LOCA and severe accident are giving a clearly picture, in how boiling heat transfer phenomenon was occurs during sequent of nuclear reactors accident, especially TMI-2 accident. The mapping of heat transfer base on transient temperature data was created in boiling curve form which was shown the differences of heat flux in three boiling regimes, both in pool boiling and flow boiling. The experimental simulation of LOCA shown that the CHF value (67.31 kW/m2 ) is small than the CHF value of severe accident (262 kW/m2). (author)

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

    International Nuclear Information System (INIS)

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

  18. Thermal and hydraulic behaviour of CANDU cores under severe accident conditions

    International Nuclear Information System (INIS)

    This report describes work performed for the Atomic Energy Control Board on a) Formation and rewetting of dry patches on CANDU reactor calandria tubes during a Loss-of-Coolant Accident, and b) Analysis of accident sequence S11: Loss-of-Coolant Accident plus Loss-of-Emergency Core Cooling plus loss of moderator cooling system. For part (a), it is concluded that any dry patches which form on calandria tubes as a result of local heating to the critical heat flux will rewet in a short time (10 to 30 seconds for a Bruce-type reactor, 90 seconds for a Douglas Point-type reactor), with negligible effects on fuel sheath and maximum pressure tube temperatures. Pressure tube integrity is not predicted to be threatened. For part (b), preliminary analysis of the S11 accident sequence is presented. The complete analysis follows in the final report on the effects of severe accidents on CANDU cores

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

    International Nuclear Information System (INIS)

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

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

    Science.gov (United States)

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

    2016-06-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-01-15

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

  3. Core/concrete interaction model for full scope simulation of severe accidents

    International Nuclear Information System (INIS)

    Nuclear plant training simulators have only recently begun to model severe loss-of-coolant accidents in which molten core material can relocate to the bottom of the reactor vessel, fail the vessel, and migrate to the containment. For those accident sequences in which core debris )corium) can accumulate in direct contact with concrete in the containment, the potential for concrete erosion and its phenomenological consequences must be assessed in order that operator training for severe accidents can be attempted. The core/concrete interaction model presented in this paper was developed for the Westinghouse full scope simulator. It allows for extension of transient simulation to conditions beyond vessel failure, and is intended for real-time operator training for severe accidents on a full scope simulator. The model predictions compare favorably with more detailed MAAP calculations

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

    Science.gov (United States)

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

    2016-03-01

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

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-06-15

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

  7. Evaluation of RCS injection strategy by normal residual heat removal system in severe accident management

    International Nuclear Information System (INIS)

    Highlights: • Integrated severe accident analysis model of ALWR RCS, ESF and containment is built. • Large-break loss of coolant accident and loss of feed water accident are analyzed. • Effectiveness of RNS injection strategy and plant system response are investigated. • Impact of RNS injection on hydrogen generation and distribution is evaluated. • Negative impact induced by different RCS depressurization measures is investigated. - Abstract: Severe Accident Management Guidelines (SAMGs) suggests mitigating the consequence of severe accident scenarios by using the non-safety systems if the safety systems are unavailable. For 1000 MWe advanced passive pressurized water reactor (PWR), the normal residual heat removal system (RNS) is proposed to implement the Reactor Coolant System (RCS) injection strategy during severe accidents if safety systems fail. Therefore, evaluation of the effectiveness and negative impact of RNS injection strategy is performed, in which two typical severe accident sequences are selected, which are the typical low-pressure core melt accident sequence induced by Large-break Loss of Coolant Accident (LLOCA) with double-ended guillotine break at cold leg and the typical high-pressure core melt accident induced by Loss of Feed Water (LOFW), to analyze RCS response using the integrated severe accident analysis code. The plant model, including RCS, Engineering Safety Features (ESF), containment and RNS, is built to evaluate the effectiveness of RNS injection by comparing the sequences with and without RCS injection, which shows that RNS injection can terminate core melt progression and maintain core cooling in these accident sequences. However, hydrogen generated during the core reflooding is investigated for the negative impact, which shows that RNS may increase the hydrogen concentration in the containment. For the sequence induced by LOFW, two different RCS depressurization measurements are compared, which shows that opening ADS

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  10. Evaluation of uncertainties in relation to severe accidents and level-2 probabilistic safety analysis

    International Nuclear Information System (INIS)

    Uncertainties of various natures have to be taken into account in severe accident analysis, in particular those related to level-2 probabilistic safety analysis (PSA). However, the extension and application of uncertainty methods to severe accidents is more difficult than for design-basis accidents because of the considerable differences in the availability of experimental data and the level of development and validation of computer codes. Best-estimate approaches used in severe accidents require an assessment of related uncertainties. Besides the evaluation of experimental data scatter, expert judgement is usually needed to assess physical parameter uncertainties, which have to be propagated to results using different techniques. Moreover, the relation between uncertainties and stochastic probabilities (concerning for instance equipment failure and human error), remains an open question, in particular in the framework of level-2 PSAs. The workshop aimed to exchange information about the state of the art in this field and to facilitate the development of a coherent approach to uncertainties in relation to severe accidents. It also provides recommendations for future NEA work in this field. These proceedings gather twenty-four articles shared into four sessions dealing with: 1 - methods for uncertainty assessment, 2 - applications to uncertainty assessment on severe accident physical phenomena, 3 - applications to uncertainty assessment in level 2 PSA, and, 4 - general discussion, conclusions and recommendations

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2011-07-01

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

  12. BWR [boiling water reactor] core criticality versus water level during an ATWS [anticipated transient without scram] event

    International Nuclear Information System (INIS)

    The BWR [boiling water reactor] emergency procedures guidelines recommend management of core water level to reduce the power generated during an anticipated transient without scram (ATWS) event. BWR power level variation has traditionally been calculated in the system codes using a 1-D [one-dimensional] 2-group neutron kinetics model to determine criticality. This methodology used also for calculating criticality of the partially covered BWR cores has, however, never been validated against data. In this paper, the power level versus water level issues in an ATWS severe accident are introduced and the accuracy of the traditional methodology is investigated by comparing with measured data. It is found that the 1-D 2-group treatment is not adequate for accurate predictions of criticality and therefore the system power level for the water level variations that may be encountered in a prototypical ATWS severe accident. It is believed that the current predictions for power level may be too high

  13. BWR 90+ - Nuclear power plant for 21st century

    International Nuclear Information System (INIS)

    BWR 90+ is a boiling water reactor, based on the previous models BWR90 and BWR75, and on the operational experiences gained with six reactors of the previous generation. The development work started in 1994 in co-operation with Teollisuuden Voima Oy (TVO). At present all the boiling water reactor owners participate the cooperation. The objectives of the development were: (1) to develop a boiling water reactor of competitive price level and short construction time, and which meets the latest safety requirements, (2) to itemize the technologies improving the security and competitivity of present plants, and (3) to maintain the expertise of the personnel of the companies participating the development work, and improving the BWR- technology. High power output and short construction time reduce the power generation costs. Large amount of fuel assemblies leads to higher safety margins. Reduction of scram groups from 18 to 16 reduces the amount of components, the assembly space and costs. The reactor technical data is as follows: Thermal power output 4250 MWth; electric power output 1500 MWe, construction time 1500 days, costs 1500 pounds/kWe, no. of fuel assemblies 872, no. of scram groups 16, turbines 1, the capacity factor 90% and the duration of service outage 3 weeks. Specific features of BWR90+ are: short construction time and low costs, risk for connection between wet and dry spaces has been minimized, reactor core remains covered by water during loss-of-coolant accident caused by fuel replacement, Passive collection and cooling of core-melt inside the containment, the containment is not the first wall against the spreading of core-melt, steam explosions and core- concrete interactions have low probabilities, high gas- volume of wet-space reduces the pressure increase during a severe accident, filter-equipped gas removal system forms the final overpressure shield, the containment is cylindrical, and the plant is equipped with digital instrumentation and control

  14. A Study On In-Vessel Severe Accident Progression In The VVER-1000 Reactor Pressure Vessel

    International Nuclear Information System (INIS)

    In a hypothetic severe accident, there is likelihood that a melt pool is formed and the Reactor Pressure Vessel (RPV) fails due to thermal creep. The present paper is concerned with the in-vessel accident progression in the VVER-1000 RPV. The study includes review of simulation results (RELAP, MELCOR), analysis of accident scenarios, determination of core materials relocation, and simulation of heat transfer of the melt pool formed in the lower plenum. The RELAP, MELCOR codes used for safety analysis are capable of simulation of severe accident progression and identification of accident scenarios in the reactor. However, there is limitation in describing turbulent natural convection heat transfer of a melt pool formed in the lower plenum. Computational Fluid Dynamics (CFD) codes which although are capable of melt pool heat transfer simulation, however, are too expensive. The Effective Convectivity Model (ECM) and Phase-change ECM (PECM) which were developed for melt pool heat transfer simulation are applied. The melt pool configuration, initial conditions are determined based on the analysis of accident scenarios, the ECM/PECM is used to simulate melt pool heat transfer. Results of ECM/PECM simulation are analyzed, compared with available RELAP, MELCOR accident progression data, RPV failure mode and timing are discussed. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    The second OECD Specialist Meeting on operator aids for severe accident management (SAMOA-2) was held in Lyon, France (1997), and was attended by 33 specialists representing ten OECD member countries. As for SAMOA-1, the scope of SAMOA-2 was limited to operator aids for accident management which were in operation or could be soon. The meeting concentrated on the management of accidents beyond the design basis, including tools which might be extended from the design basis range into the severe accident area. Relevant simulation tools for operator training were also part of the scope of the meeting. 20 papers were presented; there were two demonstrations of computerized systems (the ATLAS analysis simulator developed by GRS, and EDF's 'Simulateur Post Accidentels' (SIPA). The three sessions dealt with operator aids for control rooms, operator aids for technical support centres, and simulation tools for operator training. The various papers for each session are summarized

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

    International Nuclear Information System (INIS)

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

  20. Considerations relating to the presence of water in the reactor cavity during severe accidents

    International Nuclear Information System (INIS)

    The purpose of this paper is to present some of the factors, both positive and negative, associated with the presence of water in the reactor cavity. The presence of water in the reactor cavity is one of the factors whose influence on the evolution of severe accidents must be determined since, on the one hand, it has an impact on some of the most significant severe accident phenomena and, on the other, it could be an important factor when preparing accident management strategies resulting from containment analyses. In spite of the initial intuitive impression that water in the reactor cavity must always be beneficial, certain phenomena, such as the following must also be taken into account before developing accident management strategies: - Higher production of steam - Possibility of steam explosions - Increased production of H2 due to oxidation of steel components of the melted core ejected from the vessel - More oxidation energy released due to the presence of oxygen in the cavity (Author)