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Sample records for accident analysis code

  1. Development of criticality accident analysis code AGNES

    A one-point kinetics code, AGNES2, has been developed for the evaluation of the criticality accident of nuclear solution fuel system. The code has been evaluated through the simulation of TRACY experiments and used for the study of the condition of the JCO criticality accident. A code, AGNES-P, for the criticality accident of nuclear powder system has been developed based on AGNES2. It is expected that these codes be useful for the evaluation of criticality safety for fuel reprocessing and fabrication plants. (author)

  2. Fire-accident analysis code (FIRAC) verification

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A larger industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. We have performed a series of experiments in which a known rate of thermal energy is injected into the system. We then simulated this experiment with the FIRAC code. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

  3. Fire-accident analysis code (FIRAC) verification

    The FIRAC computer code predicts fire-induced transients in nuclear fuel cycle facility ventilation systems. FIRAC calculates simultaneously the gas-dynamic, material transport, and heat transport transients that occur in any arbitrarily connected network system subjected to a fire. The network system may include ventilation components such as filters, dampers, ducts, and blowers. These components are connected to rooms and corridors to complete the network for moving air through the facility. An experimental ventilation system has been constructed to verify FIRAC and other accident analysis codes. The design emphasizes network system characteristics and includes multiple chambers, ducts, blowers, dampers, and filters. A large industrial heater and a commercial dust feeder are used to inject thermal energy and aerosol mass. The facility is instrumented to measure volumetric flow rate, temperature, pressure, and aerosol concentration throughout the system. Aerosol release rates and mass accumulation on filters also are measured. This paper compares and discusses the gas-dynamic and heat transport data obtained from the ventilation system experiments with those predicted by the FIRAC code. The numerically predicted data generally are within 10% of the experimental data

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

    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

  5. Severe accident analysis code Sampson for impact project

    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)

  6. Severe accident analysis code Sampson for impact project

    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)

  7. The development of a severe accident analysis code

    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

  8. Severe accident analysis code SAMPSON improvement for IMPACT project

    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)

  9. Health effects estimation code development for accident consequence analysis

    As part of a computer code system for nuclear reactor accident consequence analysis, two computer codes have been developed for estimating health effects expected to occur following an accident. Health effects models used in the codes are based on the models of NUREG/CR-4214 and are revised for the Japanese population on the basis of the data from the reassessment of the radiation dosimetry and information derived from epidemiological studies on atomic bomb survivors of Hiroshima and Nagasaki. The health effects models include early and continuing effects, late somatic effects and genetic effects. The values of some model parameters are revised for early mortality. The models are modified for predicting late somatic effects such as leukemia and various kinds of cancers. The models for genetic effects are the same as those of NUREG. In order to test the performance of one of these codes, it is applied to the U.S. and Japanese populations. This paper provides descriptions of health effects models used in the two codes and gives comparisons of the mortality risks from each type of cancer for the two populations. (author)

  10. Development status of Severe Accident Analysis Code SAMPSON

    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)

  11. Development status of Severe Accident Analysis Code SAMPSON

    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)

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

    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.

  13. Adjoint-based sensitivity analysis for reactor accident codes

    This paper summarizes a recently completed study that identified and investigated the difficulties and limitations of applying first-order adjoint sensitivity methods to reactor accident codes. The work extends earlier adjoint sensitivity formulations and applications to consider problem/model discontinuities in a general fashion, provide for response (R) formulations required by reactor safety applications, and provide a scheme for accurately handling extremely time-sensitive reactor accident responses. The scheme involves partitioning (dividing) the model into submodels (with spearate defining equations and initial conditions) at the location of discontinuity. Successful partitioning moves the problem dependence on the discontinuity location from the whole model system equations to the initial conditions of the second submodel

  14. Sodium fast reactor gaps analysis of computer codes and models for accident analysis and reactor safety.

    Carbajo, Juan (Oak Ridge National Laboratory, Oak Ridge, TN); Jeong, Hae-Yong (Korea Atomic Energy Research Institute, Daejeon, Korea); Wigeland, Roald (Idaho National Laboratory, Idaho Falls, ID); Corradini, Michael (University of Wisconsin, Madison, WI); Schmidt, Rodney Cannon; Thomas, Justin (Argonne National Laboratory, Argonne, IL); Wei, Tom (Argonne National Laboratory, Argonne, IL); Sofu, Tanju (Argonne National Laboratory, Argonne, IL); Ludewig, Hans (Brookhaven National Laboratory, Upton, NY); Tobita, Yoshiharu (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Ohshima, Hiroyuki (Japan Atomic Energy Agency, Ibaraki-ken, Japan); Serre, Frederic (Centre d' %C3%94etudes nucl%C3%94eaires de Cadarache %3CU%2B2013%3E CEA, France)

    2011-06-01

    This report summarizes the results of an expert-opinion elicitation activity designed to qualitatively assess the status and capabilities of currently available computer codes and models for accident analysis and reactor safety calculations of advanced sodium fast reactors, and identify important gaps. The twelve-member panel consisted of representatives from five U.S. National Laboratories (SNL, ANL, INL, ORNL, and BNL), the University of Wisconsin, the KAERI, the JAEA, and the CEA. The major portion of this elicitation activity occurred during a two-day meeting held on Aug. 10-11, 2010 at Argonne National Laboratory. There were two primary objectives of this work: (1) Identify computer codes currently available for SFR accident analysis and reactor safety calculations; and (2) Assess the status and capability of current US computer codes to adequately model the required accident scenarios and associated phenomena, and identify important gaps. During the review, panel members identified over 60 computer codes that are currently available in the international community to perform different aspects of SFR safety analysis for various event scenarios and accident categories. A brief description of each of these codes together with references (when available) is provided. An adaptation of the Predictive Capability Maturity Model (PCMM) for computational modeling and simulation is described for use in this work. The panel's assessment of the available US codes is presented in the form of nine tables, organized into groups of three for each of three risk categories considered: anticipated operational occurrences (AOOs), design basis accidents (DBA), and beyond design basis accidents (BDBA). A set of summary conclusions are drawn from the results obtained. At the highest level, the panel judged that current US code capabilities are adequate for licensing given reasonable margins, but expressed concern that US code development activities had stagnated and that the

  15. DOE modifications to the MAAP [Modular Accident Analysis Program] code

    This report presents an enhanced model for the MAAP code that addresses fuel-cladding interaction and core mass relocation during core degradation. The main purpose of this work is to assess the potential for in-vessel hydrogen production and to reduce the uncertainty in fission product source term evaluation. The model provides a description of fuel behavior in which the fuel comprises uranium dioxide, zirconium dioxide, and U-Zr-O compounds. The composition of the U-Zr-O compounds and their solidus and liquidus temperatures are calculated throughout the core melt transient. The interaction of control rod materials with fuel and cladding and the relocation of control rod materials are not addressed in this enhanced model. The enhanced core melt progression model has been applied to a hypothetical station blackout accident with a small break via the reactor coolant pump seals. The new model has been benchmarked against both the LOFT experiment LP-FP-2 and the TMI-2 accident prior to the B-loop pump restart. Although some uncertainties and deviations were seen, general agreement was obtained with the experimental data and with the TMI-2 accident. 21 refs., 30 figs

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

    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.)

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

    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.)

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

    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.

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

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

  20. Fission product release analysis code during accident conditions of HTGR, RACPAC

    Fission product release analysis code, RACPAC (Fission Product Release Analysis Code from Fuel Particle in Accident Condition), was developed to calculate fractional release from the core during accident conditions of High Temperature Gas-cooled Reactor. RACPAC code has following features. (1) Fission product release fraction after the reactor scram is calculated based on the analytical solution with reduced diffusion coefficient. (2) The reduced diffusion coefficient for each nuclide is calculated from the (R/B) value, which is defined as release rate to birth rate of fission product. (3) The temperature transient after the accident can be taken into consideration in fractional release calculation with RACPAC. This paper describes calculation model of fission product release from fuel particle, calculation model of the reduced diffusion coefficient, users' manual and calculation examples. (author)

  1. Safety analysis of MNSR reactor during reactivity insertion accident using the validated code PARET

    In the framework of the IAEA CRP project (J7.10.10) on 'Safety significance of postulated initiating events for various types of research reactors and assessment of analytical tools' the Syrian team contributed in the assessment of computational codes related to the safety analysis of research reactors. During the project implementation the codes PARET and MERSAT have been tested, modified and verified regarding specific phenomena related to safety analysis of research reactors. In the framework of this contribution the code PARET has been applied to model the core of the Syrian MNSR reactor. The code analysis includes the simulation of steady state operation and a group of selected reactivity insertion accident (RIA) including the design basis accidents dealing with the insertion of total available excess reactivity

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

    For prevention and mitigation of the containment failure during severe accident, the study is focused on the severe accident phenomena, especially, the ones occurring inside the cavity and is intended to improve existing models and develop analytical tools for the assessment of severe accidents. A correlation equation of the flame velocity of pre mixture gas of H2/air/steam has been suggested and combustion flame characteristic was analyzed using a developed computer code. For the analysis of the expansion phase of vapor explosion, the mechanical model has been developed. The development of a debris entrainment model in a reactor cavity with captured volume has been continued to review and examine the limitation and deficiencies of the existing models. Pre-test calculation was performed to support the severe accident experiment for molten corium concrete interaction study and the crust formation process and heat transfer characteristics of the crust have been carried out. A stress analysis code was developed using finite element method for the reactor vessel lower head failure analysis. Through international program of PHEBUS-FP and participation in the software development, the research on the core degradation process and fission products release and transportation are undergoing. CONTAIN and MELCOR codes were continuously updated under the cooperation with USNRC and French developed computer codes such as ICARE2, ESCADRE, SOPHAEROS were also installed into the SUN workstation. 204 figs, 61 tabs, 87 refs. (Author)

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

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

    1995-07-01

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

  4. Model verification of the debris coolability analysis module in the severe accident analysis code 'SAMPSON'

    The debris coolability analysis module in the severe accident analysis code 'SAMPSON' has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris cooling after spreading is solved on the basis of natural convection analysis with melting and solidification on three-dimensional Cartesian co-ordinates. The calculated results for the cooling model are compared with the results from a three-dimensional natural convection experiment. The comparisons show the module capability for predictions of the debris temperature in the cooling process. Furthermore, it is seen that the prediction capability in the thermal load of the vessel wall is improved, since the penetration nozzles melting is modeled and combined with the cooling model. The module provides a good tool for the prediction of the reactor safety margin in a severe accident through the three-dimensional analysis of debris cooling. (author)

  5. Analysis code for large rupture accidents in ATR. SENHOR/FLOOD/HEATUP

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic transient change, the behavior of core reflooding and the transient change of fuel temperature in the events which are classified in large rupture accidents of reactor coolant loss, that is the safety evaluation event of the ATR, the analysis codes for thermo-hydraulic transient change at the time of large rupture SENHOR, for core reflooding characteristics FLOOD and for fuel temperature HEATUP are used, respectively. The analysis code system for loss of coolant accident comprises the analysis code for thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC in addition to the above three codes. Based on the changes with time lapse of reactor thermal output and steam drum pressure obtained by the SENHOR, average reflooding rate is analyzed by the FLOOD, and the time of starting the turnaround of fuel cladding tube temperature and the heat transfer rate after the turnaround are determined. Based on these data, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The SENHOR code, the FLOOD code and the HEATUP code and various models for these codes are explained. The example of evaluation and the sensitivity analysis of the ATR plant are reported in the Appendix. (K.I.)

  6. MELCOR code analysis of a severe accident LOCA at Peach Bottom Plant

    A design-basis loss-of-coolant accident (LOCA) concurrent with complete loss of the emergency core cooling systems (ECCSs) has been analyzed for the Peach Bottom atomic station unit 2 using the MELCOR code, version 1.8.1. The purpose of this analysis is to calculate best-estimate times for the important events of this accident sequence and best-estimate source terms. Calculated pressures and temperatures at the beginning of the transient have been compared to results from the Peach Bottom final safety analysis report (FSAR). MELCOR-calculated source terms have been compared to source terms reported in the NUREG-1465 draft

  7. EUREKA-2: a computer code for the reactivity accident analysis in a water cooled reactor

    EUREKA-2, a computer code for the reactivity accident analysis, has been developed in order to analyze neutronic, thermal and hydrodynamic transient behaviors in a water cooled reactor. EUREKA-2 can analyze the transient response of the core against the reactivity change caused by control rod withdrawal, coolant flow change and/or coolant temperature change. Especially, it can well simulate fast transient behaviors in serious reactivity accidents. This code calculates coupled neutronic and thermal-hydrodynamic responses for multi-regions in the core. EUREKA-2 has been developed by improving the fluid flow model of EUREKA and can analyze the reactivity accidents in which coolant temperature rises quickly and vapor is produced. (author)

  8. Preliminary Analysis of a Loss of Condenser Vacuum Accident Using the MARS-KS Code

    In accordance with revision of NUREG-0800 of USNRC, the area of review for loss of condenser vacuum(LOCV) accident has been expanded to analyze both peak pressures of primary and secondary system separately. Currently, the analysis of LOCV accident, which is caused by malfunction of condenser, has been focused to fuel cladding integrity and peak pressure in the primary system. In this paper, accident analysis for LOCV using MARS-KS code were conducted to support the licensing review on transient behavior of secondary system pressure of APR1400 plant. The accident analysis for the loss of condenser vacuum (LOCV) of APR1400 was conducted with the MARS-KS code to support the review on the pressure behavior of primary and secondary system. Total four cases which have different combination of availability of offsite power and the pressurizer spray are considered. The preliminary analysis results shows that the initial conditions or assumptions which concludes the severe consequence are different for each viewpoint, and in some cases, it could be confront with each viewpoint. Therefore, with regard to the each acceptance criteria, figuring out and sensitivity analysis of the initial conditions and assumptions for system pressure would be necessary

  9. Preliminary Analysis of a Loss of Condenser Vacuum Accident Using the MARS-KS Code

    Kim, Jieun Kim; Bang, Young Seok; Oh, Deog Yeon; Kim, Kap; Woo, Sweng-Wong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-05-15

    In accordance with revision of NUREG-0800 of USNRC, the area of review for loss of condenser vacuum(LOCV) accident has been expanded to analyze both peak pressures of primary and secondary system separately. Currently, the analysis of LOCV accident, which is caused by malfunction of condenser, has been focused to fuel cladding integrity and peak pressure in the primary system. In this paper, accident analysis for LOCV using MARS-KS code were conducted to support the licensing review on transient behavior of secondary system pressure of APR1400 plant. The accident analysis for the loss of condenser vacuum (LOCV) of APR1400 was conducted with the MARS-KS code to support the review on the pressure behavior of primary and secondary system. Total four cases which have different combination of availability of offsite power and the pressurizer spray are considered. The preliminary analysis results shows that the initial conditions or assumptions which concludes the severe consequence are different for each viewpoint, and in some cases, it could be confront with each viewpoint. Therefore, with regard to the each acceptance criteria, figuring out and sensitivity analysis of the initial conditions and assumptions for system pressure would be necessary.

  10. A simulation of steam generator tube rupture accident by safety analysis code RELAP5/MODI

    Steam-generator-tube-rupture accident occurred at Prairie Island unit 1 is simulated using the RELAP5/MOD1 code which has been developed as a best-estimate safety analysis code for light water reactors. The purpose of the simulation is to examine its capacity as a tool of obtaining high-quality and verified data base needed for developing diagnostic techniques of nuclear power plants. The simulation is conducted until 3200 seconds after the tube rupture. The simulation results agrees fairly well with both the plant records and the RETRAN-02 simulation results conducted at Japan Atomic Energy Research Institute, and it is concluded that the RELAP5/MOD1 code is effective to simulate the overall plant behavior during the accident, although several items remain for future improvement. (author)

  11. Analysis of TRACY experiment and JCO criticality accident by using AGNES code

    A one-point kinetics code, AGNES, has been developed in JAERI for the purpose of the analysis of TRACY experiment. Four of the experiments performed in ramp feed mode were simulated by AGNES code, and the power, temperature and total fission number were evaluated. The calculated values of them were in agreement with the experimental values with ±15% error. In the analysis of JCO criticality accident, three supposed cases were considered, and the total fission number was evaluated at 4 - 6x1017 by insertion of 1.5 - 3.0$ excess reactivity. (author)

  12. Application of MELCOR code to the MCCI analysis in Severe Accident Sequences

    This paper provides some of the technical aspects that can be applied to an analysis of the MCCI phenomena in a severe accident scenario using the current MELCOR version. An application methodology of the MELCOR current version to the analysis of MCCI, the phenomena of which are very uncertain and lack specific knowledge during severe accidents, was introduced. Assumptions based on the experimental results are used instead of the phenomenological detail modeling because of the modeling limitations. In the technical aspects of MCCI, code modification itself is not a big deal, because the code modification is needed for just the user flexibility. The concern will be whether the assumptions made for this analysis are acceptable or not. This paper illustrates the application of a severe accident analysis code, MELCOR, to the analysis of molten corium-concrete interaction (MCCI) phenomena in cases of severe accidents in nuclear power plants. In postulated degraded core accidents, followed by the failure of certain engineered safety features of the reactor system, the reactor core may eventually melt owing to the generation of decay heat. If the safety features of the reactor system fail to arrest the accident within the reactor vessel, the corium (molten core debris) will fall into the reactor cavity and attack the concrete walls and floor. Basemat melt-through refers to the process of concrete decomposition and destruction associated with a corium melt interacting with the reactor cavity basemat. The potential hazard of MCCI is the integrity of the containment building owing to the possibility of a basemat melt-through, containment overpressurization by non-condensible gases, or the oxidation of combustible gases. In the meantime, the MCCI still has large uncertainties in several phenomena such as melt spreading area, debris particulation, and heat transfer between the debris and cooling water. In particular, in the case where the water pool exists in the reactor

  13. Coremelt-2D Code for Analysis of Severe Accidents in a Sodium Fast Reactor

    In the paper there is a description of COREMELT-2D code designed for carrying out coupled two-dimensional analysis of neutronic and thermohydraulic transients, which may occur in the core of sodium cooled fast reactor (SFR), including severe accidents resulting in damage of SFR core and relocation of its components with the change of their aggregative state, namely: boiling and condensation of coolant, damage and melting of fuel element claddings and fuel, relocation of molten core components, thermal interaction of fuel and coolant and freezing of steel and fuel. So, COREMELT-2D code is capable of analyzing all stages of ULOF accident up to expansion phase characterized by the intensive interaction of molten fuel and sodium. Modular structure of COREMELT-2D code consisting of thermohydraulic module COREMELT and neutronic module RADAR is presented. Preservation equations are solved in COREMELT module in two-dimensional cylindrical R-Z geometry in porous body approximation. RADAR module is used for solving multi-group neutron diffusion equation in R-Z and X-Y geometry. Application of the code for solving dynamics tasks with rather rapid changes of neutron constants requires efficient unit for constants preparation. For this purpose, steady state analysis TRIGEX code (HEX-Z geometry) is used, which includes the program of nuclear data preparation CONSYST connected to the ABBN-93 group constants library. In the paper presented are the results of comparative analytical studies on ULOF beyond design severe accident as applied to the BN-1200 reactor design made by COREMELT-2D code and by its previous version based on neutron kinetics point model. The results of analysis make it possible to evaluate the effect of space-time changes of reactor neutronics caused by sodium removal from the core as a result of sodium boiling. (author)

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

    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.

  15. RAVE code system for 3-D core non-LOCA accident analysis

    Full text of publication follows: This paper provides an overview of the application of the Westinghouse updated RAVE three dimensional (3-D) core transient analysis code system for PWR non-LOCA accident analysis. The RAVE code system consists of a linkage of the following USNRC-approved codes: the EPRI RETRAN-02 (RETRAN) system transient analysis code, the Westinghouse SPNOVA (also referred to as ANC-K) reactor core neutron kinetic nodal code, and the EPRI VIPRE-01 (VIPRE) reactor core thermal-hydraulic (T/H) code. The RETRAN code is used for calculating transient conditions in the reactor coolant system (RCS), including reactor vessel, RCS loops, pressurizer and steam generators. RETRAN also models reactor trips, engineering safety feature (ESF) functions, and the control systems. The SPNOVA code is used to perform 3-D core neutronic calculations for core average power and power distributions in the core. Its reactivity feedback calculation is based on transient fluid conditions and fuel temperatures obtained from the VIPRE code. Based on core inlet temperature, inlet flow and core exit pressure from RETRAN, and the nodal nuclear power from SPNOVA, VIPRE provides back to RETRAN transient nodal heat flux in the reactor core region. An effective 3-D analysis requires RETRAN, SPNOVA and VIPRE calculations to be closely linked for the entire reactor core. The linking architecture uses a standard external communication interface protocol for communication among the running programs on the same or different computers. The RAVE code system currently uses the Parallel Virtual Machine (PVM) software for the data transfer. Besides the necessary changes for data transfer, no other changes were made to RETRAN, SPNOVA or VIPRE fundamental code algorithms or solution methods. The RETRAN model in the RAVE system uses the same detailed reactor vessel, RCS loops, pressurizer, and steam generator, and control and protection models as has been licensed for current plant Safety

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

    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.

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

    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

  18. Code portability and data management considerations in the SAS3D LMFBR accident-analysis code

    The SAS3D code was produced from a predecessor in order to reduce or eliminate interrelated problems in the areas of code portability, the large size of the code, inflexibility in the use of memory and the size of cases that can be run, code maintenance, and running speed. Many conventional solutions, such as variable dimensioning, disk storage, virtual memory, and existing code-maintenance utilities were not feasible or did not help in this case. A new data management scheme was developed, coding standards and procedures were adopted, special machine-dependent routines were written, and a portable source code processing code was written. The resulting code is quite portable, quite flexible in the use of memory and the size of cases that can be run, much easier to maintain, and faster running. SAS3D is still a large, long running code that only runs well if sufficient main memory is available

  19. Fuel Behavior Simulation Code FEMAXI-FBR Development for SFR Core Disruptive Accident Analysis

    Japan Nuclear Energy Safety Organization (JNES) has been developing ASTERIA-FBR code system for SFR core disruptive accident analysis to contribute as a part of the regulation activity for Japanese prototype FBR, MONJU. The ASTERIA-FBR code system consists of detailed fuel behavior analysis module (FEMAXI-FBR), neutronic Monte-Carlo calculation module (GMVP), and thermal hydraulic module (CONCORD). The calculation scope of the ASTERIA-FBR covers the initiating, transitional and post disassembly expansion processes. The FEMAXI-FBR is based on LWR fuel behavior simulation code FEMAXI-6 and modified the material properties and the calculation models under steady state and transient operational condition. The FEMAXI-FBR has been verified in steady state calculations compared with those of SAS-4A code. Furthermore, the code has been validated by French CABRI slow-TOP (E12) and fast-TOP (BI2) transient calculations. Through these verification and validation, good agreement has been obtained with the FP-gas release ratio, the fuel restructuring, the gap width between pellet and cladding, and the fuel pin failure position. (author)

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

    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)

  1. Analysis code for medium and small rupture accidents in ATR. LOTRAC/HEATUP

    NONE

    1997-08-01

    In the evaluation of thermo-hydraulic and fuel temperature transient changes in the events which are classified in medium and small rupture accidents of reactor coolant loss that is the safety evaluation event of the ATR, the analysis code for synthetic thermo-hydraulic transient change at the time of medium and small ruptures LOTRAC and the detailed analysis code for fuel temperature HEATUP are used, respectively. By using the LOTAC, the thermo-hydraulic behavior of reactor cooling facility and the temperature behavior of fuel at the time of blow-down are analyzed, and also the characteristics of changing reactor thermal output is analyzed, considering the functioning characteristics of emergency core cooling system. Based on the data of thermo-hydraulic behavior obtained by the LOTRAC, the time of beginning the turn-around of fuel cladding tube temperature obtained by the data of ECCS pouring characteristics, the heat transfer rate after the turn-around and so on, the detailed temperature change of fuel elements is analyzed by the HEATUP, and the highest temperature and the amount of oxidation of fuel cladding tubes are determined. The LOTRAC code, the HEATUP code, various analysis models, and rupture simulation experiment are reported. (K.I.)

  2. Model Development of Light Water Reactor Fuel Analysis Code RANNS for Reactivity-initiated Accident Conditions

    A light water reactor fuel analysis code RANNS has been developed to analyze thermal and mechanical behaviors of a single fuel rod in mainly Reactivity-Initiated Accident (RIA) conditions, based on the light water reactor fuel analysis code FEMAXI-7, which has been developed for normal operation conditions and anticipated transient conditions. The recent model development for the RANNS code has been focused on improving predictability of stress, strain, and temperature inside a fuel rod during pellet cladding mechanical interaction (PCMI), which is one of the most important behaviors of high-burnup fuels under RIA conditions. This report provides descriptions of the models developed and/or validated recently via experimental analyses using the RANNS code on the RIA-simulating experiments conducted in the Nuclear Safety Research Reactor (NSRR): models for mechanical behaviors as relocation of fuel pellets, pellet yielding, pellet-cladding mechanical bonding, and PCMI failure limit of fuel cladding, and thermal behaviors as pellet-cladding gap conductance and heat transfer from fuel rod surface to coolant water. (author)

  3. Accident consequence assessment code development

    This paper describes the new computer code system, OSCAAR developed for off-site consequence assessment of a potential nuclear accident. OSCAAR consists of several modules which have modeling capabilities in atmospheric transport, foodchain transport, dosimetry, emergency response and radiological health effects. The major modules of the consequence assessment code are described, highlighting the validation and verification of the models. (author)

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

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

    2014-10-15

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

  5. Analysis of Fukushima Daiichi Nuclear Power Plant by SAMPSON severe accident code - Unit 3

    On March 11th 2011 an extremely high magnitude earthquake and following tsunami struck the East coast of Japan, resulting in a nuclear accident unprecedented in time and extent. After scram started at all power stations, diesel generators began operation until tsunami waves reached the power plants. Flooding by tsunami had a great impact on the plant safety systems availability, leading to the condition of station black out at Fukushima Daiichi from unit 1 to 3. In the present work the severe accident code SAMPSON is employed for the analysis of the first part of the transient in Fukushima Daiichi Unit 3. In this unit DC batteries remained available for about 40 hours after scram and influenced the time of melting onset, hydrogen release in the reactor building and further explosion. Models of high pressure safety systems were improved in SAMPSON, considering a more realistic pump-turbine unit operation and communication between reactor and containment. Moreover new suppression pool stratification and spray models were developed and implemented in the containment module, showing a great impact in the drywell pressure estimation. It has been shown how the computed results of pressure (in reactor vessel and drywell) and core water level show a fair agreement with the measurement data and notably improvements compared to the previous analyses. (author)

  6. Overview of the IMPACT severe accident analysis code SAMPSON - On core degradation and lower plenum debris behavior in the main

    The first version of the IMPACT-SAMPSON code was completed. SAMPSON is the best estimate integral code for severe accident analysis with modular structure. Each module can run independently and communication with multiple analysis modules supervised by the analysis control module makes an integral analysis possible while appearing to users to be a single code. Multi-dimensional mechanistic models and theoretical-base equations were applied. An execution of enormous amount of calculation steps becomes possible with the use of a parallel processing computer. Models in each module were verified by test analyses. Final integral verification by PHEBUS test analyses is in progress and integral analyses of light water nuclear power plants will be performed to demonstrate quantitatively that adequate safety margin exists to cope with severe accidents. (authors)

  7. Analysis of verification and validation problems of calculation means (codes) of accident thermohydrodynamic processes for domestic NPPs

    Analysis of known approaches in area of verification and validation of calculation means (codes) modelling the accident / transition processes in nuclear power plant (NPP) equipment is represented in this review article. Needs to develop and realise the generalised calculation means verification / validation methodology taking into account, together with traditional procedures, the codes applicability assessment criteria for decision of specific tasks and for specific equipment, mathematical models and experimental stands adequacy to full-scale conditions are shown

  8. Development of debris coolability analysis module in severe accident analysis code SAMPSON for IMPACT project

    Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability. (author)

  9. Development of debris coolability analysis module in severe accident analysis code SAMPSON for IMPACT project

    Ujita, Hiroshi [Advanced Simulation Systems Department, Nuclear Power Engineering Cooperation, Tokyo (Japan); Hidaka, Masataka; Susuki, Akira; Ishida, Naoyuki

    1999-10-01

    Debris coolability in the lower plenum of the reactor pressure vessel is an important factor for the evaluation of in-vessel debris retention. The debris coolability analysis module has been developed to predict more mechanistically the safety margin of the present reactor vessels in a severe accident. The module calculates debris spreading and cooling through melting and solidification in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the explicit method on a quasi-three-dimensional scheme and debris coolability is solved on the basis of natural convection analysis with melting and solidification. The calculated results for spreading were compared with the results from a water spreading experiment on the floor and the results for coolability were compared with those from an n-octadecane melting experiment in the rectangular vessel. The comparisons showed the capability for predictions of the spearhead transportation in the debris spreading process and of the melting front transportation and time evolution of the fluid temperature in the melting process. The module provides a good tool for the prediction of the reactor pressure vessel safety margin in a severe accident through the analysis of debris spreading and coolability. (author)

  10. Development of GRIF-SM: The code for analysis of beyond design basis accidents in sodium cooled reactors

    GRIF-SM code was developed at the IPPE fast reactor department in 1992 for the analysis of transients in sodium cooled fast reactors under severe accident conditions. This code provides solution of transient hydrodynamics and heat transfer equations taking into account possibility of coolant boiling, fuel and steel melting, reactor kinetics and reactivity feedback due to variations of the core components temperature, density and dimensions. As a result of calculation, transient distribution of the coolant velocity and density was determined as well as temperatures of the fuel pins, reactor core and primary circuit as a whole. Development of the code during further 6 years period was aimed at the modification of the models describing thermal hydraulic characteristics of the reactor, and in particular in detailed description of the sodium boiling process. The GRIF-SM code was carefully validated against FZK experimental data on steady state sodium boiling in the electrically heated tube; transient sodium boiling in the 7-pin bundle; transient sodium boiling in the 37-pin bundle under flow redaction simulating ULOF accident. To show the code capabilities some results of code application for beyond design basis accident analysis on BN-800-type reactor are presented. (author)

  11. Accident analysis of Fukushima Daiichi NPP Unit-1 with SAMPSON code

    The progress of the core disruption of the Fukushima Daiichi NPP Unit-1 was analyzed by the severe accident analysis code SAMPSON. The code includes new modellings of the phenomena that occurred which have been deemed specific to the Fukushima Daiichi NPP: (1) steam leakage from the gasket of the safety relief valve (SRV) and from the buckling portion of the guide tubes (GTs) of some in-core monitors (source range monitors (SRMs) and intermediate range monitors (IRMs)): (2) melting of SRM/IRM GTs at the bottom of the reactor pressure vessel (RPV); and (3) incorporation of continuous drainage pathways for debris relocation. During the early phase of the accident after the reactor scram, the isolation condensers (ICs) had intermittently worked until the loss of AC and DC power supplies by the tsunami. The analysis reproduced well the RPV pressure transient during the IC operation period. After the loss of AC and DC power supplies, the SRV had repeated its opening and closing to keep the RPV pressure constant at about 7.5 MPa for about 4.5 hours, resulting in a gradual decrease of water level in the core. Then the SRV stopped working due to depressurization by the direct steam release from the buckling portions of the SRM/IRM GTs and from the SRV gasket. The eutectic B4C (control rod material) and steel reacted, resulting in the initiation of melting at about 4.5 h after the scram when the collapsed water level was getting closer to the bottom of active fuel, followed by melting of steel, zircalloy, and eutectics of UO2+Zr. All 12 SRM/IRM GTs had sequentially melted at about 6.5 h after the scram, resulting in fall down of melts onto the pedestal floor. Since there was no intentional core cooling for about 14 hours after the termination of the ICs until the alternative water injection by a fire engine, the core disruption continued. When the alternative water injection was started at 05:46, March 12 (15 h after the scram), 85% of the core materials had already become

  12. Thermal hydraulic studies of undercooling accidents in LMFBR safety analysis: Codes and validation

    This communication is related to the LMFBR safety analysis of undercooling accidents such as pump run down or total inlet blockage of a subassembly. The authors present the physical models developed for sodium boiling propagation and clad motion and their application to SCARABEE in pile experiments simulating loss of flow accidents in bundle geometry. These studies showed the validity of our description of boiling propagation and improved our understanding of the clad relocation phenomena

  13. Results of a survey on accident and safety analysis codes, benchmarks, verification and validation methods

    This report is a compilation of the information submitted by AECL, CIAE, JAERI, ORNL and Siemens in response to a need identified at the 'Workshop on R and D Needs' at the IGORR-3 meeting. The survey compiled information on the national standards applied to the Safety Quality Assurance (SQA) programs undertaken by the participants. Information was assembled for the computer codes and nuclear data libraries used in accident and safety analyses for research reactors and the methods used to verify and validate the codes and libraries. Although the survey was not comprehensive, it provides a basis for exchanging information of common interest to the research reactor community

  14. SACO-1: a fast-running LMFBR accident-analysis code

    Mueller, C.J.; Cahalan, J.E.; Vaurio, J.K.

    1980-01-01

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed.

  15. SACO-1: a fast-running LMFBR accident-analysis code

    SACO is a fast-running computer code that simulates hypothetical accidents in liquid-metal fast breeder reactors to the point of permanent subcriticality or to the initiation of a prompt-critical excursion. In the tradition of the SAS codes, each subassembly is modeled by a representative fuel pin with three distinct axial regions to simulate the blanket and core regions. However, analytic and integral models are used wherever possible to cut down the computing time and storage requirements. The physical models and basic equations are described in detail. Comparisons of SACO results to analogous SAS3D results comprise the qualifications of SACO and are illustrated and discussed

  16. Accident analysis in the water loop of the nuclear engineering department of IPEN using the RELAP4 code

    A thermal-hydraulic analysis to describe the transient behavior in the water loop of the Nuclear Engineering Department of the Instituto de Pesquisas Energeticas e Nucleares, Sao Paulo, Brazil, was performed. Postulated accidents such as those resulting from (1) loss of coolant, (2) main pump failure and (3) power excursions, were studied. The computer code RELAP4/Mod.3 was employed as the principal tool of analysis. (Author)

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

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

  18. JERICHO computer code: PWR containment response during severe accidents description and sensitivity analysis

    The JERICHO code has been developed in order to study the thermodynamic behaviour inside the reactor containment building for the complete spectrum of accident sequences likely to occur in such a reactor, including models for the various mass and energy transfer phenomena, for water spray, for hydrogen and carbon monoxide flammability limits and combustion, as well as for containment venting. Sensitivity analyses have been performed on a severe accident sequence, (namely, small LOCA with failure of the emergency core cooling and containment spray systems), involving core melting and subsequent concrete containment basemat erosion. The effect of various models, such as mass and energy transfer to the structures, has been studied. The influence of the concrete composition, of the fission product deposition and of the thermal degradation of the reactor cavity concrete walls on long term thermodynamic behaviour has also been investigated

  19. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  20. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

    This paper presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is in an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  1. Accident analysis of TEPCO's Fukushima Daiichi Nuclear Power Plant with the SAMPSON severe accident code. (2) Unit 1 analysis with improved debris relocation model

    On March 11, 2011, the Great Eastern Japan earthquake and the subsequent tsunami caused the station black out at TEPCO’s Fukushima Daiichi Nuclear Power Plants, and the events that followed led to core meltdowns. For assessment of the present core status, simulations have been performed with the SAMPSON severe accident code. The core debris relocation behaviors are newly investigated in this paper by applying the improved debris relocation model to the analysis of the Fukushima Daiichi unit 1 with SAMPSON code. The improvements to the model are as follows. (1) The velocity limiters and control rod guide tubes are newly taken into account. (2) The flow path of debris is modified so that it goes directly down to the lower plenum through the orifice, while in the old model, the debris had stayed on the core plate until the plate melted. In the plant analysis of unit 1 with the improved model, more than 96 wt% of the core debris is particulate. Much of debris, mainly composed of the fuel and zirconium particle, goes out of the core region through the orifice, while the debris falling on the velocity limiters is mainly composed of steel and control rod material particles. (author)

  2. Status on development and verification of reactivity initiated accident analysis code for PWR (NODAL3)

    A coupled neutronics thermal-hydraulics code NODAL3 has been developed based on the nodal few-group neutron diffusion theory in 3-dimensional Cartesian geometry for a typical pressurized water reactor (PWR) static and transient analyses, especially for reactivity initiated accidents (RIA). The spatial variables are treated by using a polynomial nodal method (PNM) while for the neutron dynamic solver the adiabatic and improved quasi-static methods are adopted. A simple single channel thermal-hydraulics module and its steam table is implemented into the code. Verification works on static and transient benchmarks are being conducting to assess the accuracy of the code. For the static benchmark verification, the IAEA-2D, IAEA-3D, BIBLIS and KOEBERG light water reactor (LWR) benchmark problems were selected, while for the transient benchmark verification, the OECD NEACRP 3-D LWR Core Transient Benchmark and NEA-NSC 3-D/1-D PWR Core Transient Benchmark (Uncontrolled Withdrawal of Control Rods at Zero Power). Excellent agreement of the NODAL3 results with the reference solutions and other validated nodal codes was confirmed. (author)

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

    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)

  4. Uncertainty analysis for control rod ejection accidents simulated by KIKO3D/TRABCO code system

    Recently, considerable conservatism must be applied in the traditional safety analyses for taking into account the uncertainties originating from the input parameters, approximations in the models, due to the safety reserves, etc. The extreme values for all of the input parameters are supposed in the traditional safety analysis at the same time. Additionally it must be mentioned that the selection of the input parameter values leading to conservative results often is not easy. The main goal of this paper is to present a more realistic methodology for the case of control rod ejection accidents. The applied consistent statistical approach leads to conservative results also, but avoids the unnecessary cumulative conservatism. A method based on a mathematical model ('Two-Sided Statistical Tolerance Intervals', [1-2]) was chosen for the realization of uncertainty analyses of Reactivity Initiated Accidents (RIA). (author)

  5. Development of severe accident Analysis Code SAMPSON in super simulator IMPACT' project

    Morii, Tadashi; Ujita, Hiroshi; Vierow, Karen; Naitoh, Masanori [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan); Yamagishi, Makoto

    1999-07-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology', project Phase 1 have been completed. At the end of Phase 1, the Basic Single-, Two-, Multi-Phase Flow Analysis Modules of Various Coordinates have been parallelized. The physical models in the Boiling Transition Analysis Code and the Fluid-Structure Interaction Analysis Code have been completed and verified by comparison with basic experimental results. The verification study of the code was conducted in two steps. First, each analysis module was run independently and analysis results were compared against separate-effect experiment data. Verification analyses included: 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. All comparison showed good agreement. Second, with the Simulation Supervisory Module, these analysis modules were executed concurrently in the parallel environment to demonstrate the code capability and integrity. (J.P.N.)

  6. Severe accident analysis methodology in support of accident management

    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

  7. Qualification of the ARROTTA code for light water reactor accident analysis

    Qualification efforts have been performed by the Taiwan Power Company (TPC) and the Institute of Nuclear Energy Research (INER) for the three-dimensional spatial kinetics code ARROTTA for light water reactor (LWR) core transient analysis. Together TPC and INER started a 5-yr project in 1989 to establish independent capabilities to perform reload design and transient analysis utilizing state-of-the-art computer programs. As part of the effort, the ARROTTA code was chosen to perform multidimensional kinetics calculations such as rod ejection for pressurized water reactors and rod drop for boiling water reactors (BWR). To qualify ARROTTA for evaluation of the Final Safety Analysis Report licensing basis core transients, ARROTTA has been benchmarked for the static core analysis against plant measured data and SIMULATE-3 predictions, and for the kinetic analysis against available benchmark problems. The static calculations compared include critical boron concentration, core power distribution, and control rod worth. The results indicate that ARROTTA predictions match very well with plant measured data and SIMULATE-3 predictions. The kinetic benchmark problems validated include the Nuclear Energy Agency Committee on Reactor Physics rod ejection problem, the three-dimensional Langenbuch-Maurer-Werner LWR rod withdrawal/insertion problem, and the three-dimensional linear regression analysis BWR transient benchmark problem. The results indicate that ARROTTA's accuracy and stability are excellent as compared with other space-time kinetics codes. It is therefore concluded that ARROTTA provides accurate predictions for multidimensional core transients for LWRs

  8. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    Abd, Aziz Sadri [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of); Shin, Andong [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident.

  9. Loss of Coolant Accident Analysis for 1MW PUSPATI Triga Mark II Research Reactor (RTP) Using MARS-KS Code

    RTP is a pool type reactor cooled by natural circulation and the reactor core is located at the bottom of a demineralized water-filled aluminum liner tank of 2.0 meter diameter and 6.5 meter depth. The core assembly is composed of 100 cylindrical fuel rods including of 4 control rods in circular array. From the literature, development of thermal hydraulic analysis of RTP using computer code has not been well established. Therefore, establishment and development of appropriate thermal hydraulic safety analysis model is very critical to ensure the safety operation of the reactor. Hence, key thermal hydraulic parameters of RTP reactor operating under steady state and transient condition were investigated. In this paper, Loss Of Coolant Accident (LOCA) were calculated and analyzed and compared with corresponding values in Safety Analysis Report (SAR) 2008 and test report. PUSPATI Triga Mark II research reactor (RTP) has been operated at Malaysian Nuclear Agency since 1982 and primary cooling system was modified in 2010. Thermal hydraulic modeling of RTP of 1MWt has been successfully investigated with MARS-KS code. The calculated normal operation parameters have been compared with reactor Safety Analysis Report (SAR) and experimental data. Most of the thermal hydraulic parameters show good agreement with SAR and experimental data within an acceptable percentage error. The loss of coolant accident was simulated in case of leak of primary side heat exchanger gasket. The calculation result showed fast decrease of reactor pool level. About 5 minutes after the leak, reactor tank was fully depleted. Furthermore, claddings temperature was reached 1173.4K at 3270s which could result in failure of SS304 cladding. Based on the assessment, it is found that appropriate remedies including physical modifications or emergency procedures need be prepared to protect the reactor tank depletion by the heat exchanger leak accident

  10. Accident analysis of flow blockage to coolant channels of upgraded JRR-3, using EUREKA-2 code, (1)

    This report describes the results about thermo-hydraulic behavior in the accident of flow blockage to coolant channels of upgraded JRR-3. Analysis was carried out using EUREKA-2 code. Flow blockage to coolant channels accident occur by some extraneous things which come from outside of the reactor pool, may block the coolant flow channels of the core. If flow blockage to coolant channels would occur, fuel temperature will increase due to flow rate decrease of coolant channels. And at last, fission products will be released from inside of fuel plates to the primary cooling system due to failure of fuel plates. In the analysis, one standard type fuel element was supposed as flow blockage channels, in the same way sa one of credible accidents, which postulated in the JRR-3 safety assessment. From the results, it was shown that about 16.7 % of the fuel element which was supposed as flow blockage channels, would fail, assuming that fuel plates might fail when the fuel meat temperatures riseover 400 deg C. (author)

  11. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    Pavlovitchev, A.M.

    2000-03-08

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes.

  12. Kinetics Parameters of VVER-1000 Core with 3 MOX Lead Test Assemblies To Be Used for Accident Analysis Codes

    The present work is a part of Joint U.S./Russian Project with Weapons-Grade Plutonium Disposition in VVER Reactor and presents the neutronics calculations of kinetics parameters of VVER-1000 core with 3 introduced MOX LTAs. MOX LTA design has been studied in [1] for two options of MOX LTA: 100% plutonium and of ''island'' type. As a result, zoning i.e. fissile plutonium enrichments in different plutonium zones, has been defined. VVER-1000 core with 3 introduced MOX LTAs of chosen design has been calculated in [2]. In present work, the neutronics data for transient analysis codes (RELAP [3]) has been obtained using the codes chain of RRC ''Kurchatov Institute'' [5] that is to be used for exploitation neutronics calculations of VVER. Nowadays the 3D assembly-by-assembly code BIPR-7A and 2D pin-by-pin code PERMAK-A, both with the neutronics constants prepared by the cell code TVS-M, are the base elements of this chain. It should be reminded that in [6] TVS-M was used only for the constants calculations of MOX FAs. In current calculations the code TVS-M has been used both for UOX and MOX fuel constants. Besides, the volume of presented information has been increased and additional explications have been included. The results for the reference uranium core [4] are presented in Chapter 2. The results for the core with 3 MOX LTAs are presented in Chapter 3. The conservatism that is connected with neutronics parameters and that must be taken into account during transient analysis calculations, is discussed in Chapter 4. The conservative parameters values are considered to be used in 1-point core kinetics models of accident analysis codes

  13. Integrated verification test of Severe Accident Analysis Code SAMPSON in super Simulation 'IMPACT' system

    Ujita, Hiroshi; Naitoh, Masanori [Advanced Simulation Systems Department, Nuclear Power Engineering Corporation, Tokyo (Japan); Karasawa, Hidetoshi; Miyagi, Kazumi

    1999-07-01

    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. 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 analysed and phenomena occurred in scenarios can be simulated quantitatively reasonably considering the physical models used for the situation. (author)

  14. MELCOR Accident Consequence Code System (MACCS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs

  15. MELCOR Accident Consequence Code System (MACCS)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  16. MELCOR Accident Consequence Code System (MACCS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projections, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management

  17. Qualification of the WIMS lattice code, for the design, operation and accident analysis of nuclear reactors

    A basic problem in nuclear reactor physics in that of the description of the neutron population behaviour in the multiplicative medium of a nuclear fuel. Due to the magnitude of the physical problem involved and the present degree of technological evolution regarding computing resources, of increasing complexity and possibilities, the calculation programs or codes have turned to be a basic auxiliary tool in reactor physics. In order to analyze the global problem, several aspects should be taken into consideration. The first aspect to be considered is that of the availability of the necessary nuclear data. The second one is the existence of a variety of methods and models to perform the calculations. The final phase for this kind of analysis is the qualification of the computing programs to be used, i.e. the verification of the validity domain of its nuclear data and the models involved. The last one is an essential phase, and in order to carry it on great variety of calculations are required, that will check the different aspects contained in the code. We here analyze the most important physical processes that take place in a nuclear reactor cell, and we consider the qualification of the lattice code WIMS, that calculates the neutronic parameters associated with such processes. Particular emphasis has been put in the application to natural uranium fuelled reactor, heavy water cooled and moderated, as the Argentinean power reactors now in operation. A wide set of experiments has been chosen: a.-Fresh fuel in zero-power experimental facilities and power reactors; b.-Irradiated fuel in both types of facilities; c.-Benchmark (prototype) experiments with loss of coolant. From the whole analysis it was concluded that for the research reactors, as well as for the heavy water moderated power reactors presently operating in our country, or those that could operate in a near future, the lattice code WIMS is reliable and produces results within the experimental values and

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

    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

  19. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  20. Thermal-hydraulic system analysis using the MARS code for the transient steam generator tube rupture accident

    A postulated SGTR accident of the APR1400 was analysed using the best estimate safety analysis code, MARS. The main objective of this study is not only to provide physical insight into the system response of the APR1400 reactor during a SGTR but also to investigate the effect of reactor trip type of a HSGL and a LPP on the thermal-hydraulic system response. As for the tube rupture modelling method, double tube modelling was adopted. Broken U-tubes were modelled as a separate assembly of a single volume. The reactor trip type affected the overall progress of the major events. However, the effect on the thermal-hydraulic response of the plant was trivial. (author)

  1. Analysis of transients and accidents with the system code ATHLET for the Krsko Nuclear Power Plant

    Main aspects of the cooperation between the republic of Croatia and the F.R. of Germany in the field of NPP safety research are overviewed in the paper. The GRS system code ATHLET developed for the analysis of anticipated and abnormal plant transients, small and intermediate leaks as well as large breaks in light water reactors is now being available at the University of Zagreb. A very comprehensive ATHLET standard input data set for the NPP Krsko has been established. This data set was validated by calculation of the event 'Main Steam Isolation Valve Closure' that occurred at the PWR NPP Krsko in 1995 and comparing the resulting characteristic parameters with the corresponding measured data. (author)

  2. Aerosol transport analysis of LWR high-consequence accidents using the HAA-4A code

    Use of the HAA-4A code to calculate removal of aerosol in containment due to inherent behavior mechanisms is described. Results for a PWR TMLB' scenario showed a source reduction of about a factor of 50 in CsI available for release to the environment through a catastrophic containment failure. Respirable CsI entering containment from the primary coolant system and melt-through blowdown was a factor of 25 less than the source. The principal removal mechanisms were particle growth due to Brownian and differential settling agglomeration and subsequent fallout. Sensitivities to important and uncertain parameters are discussed. Increased removal due to turbulent agglomeration and a larger expected source particle size are indicated. A seven control volume analysis took less than 1 minute of CPU time on an IBM 3033

  3. MELCOR Accident Consequence Code System (MACCS)

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems

  4. MELCOR Accident Consequence Code System (MACCS)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  5. Linking of FRAP-T, FRAPCON and RELAP-4 codes for transient analysis and accidents of light water reactors fuel rods

    The computer codes FRAP-T, FRAPCON and RELAP-4 have been linked for the fuel rod behavior analysis under transients and hypothetical accidents in light water reactors. The results calculated by thermal hydraulic code RELAP-4 give input in file format into the transient fuel analysis code FRAP-T. If the effect of fuel burnup is taken into account, the fuel performance code FRAPCON should provide the initial steady state data for thhe transient analysis. With the thermal hydraulic boundary conditions provided by RELAP-4 (MOD3), FRAP-T6 is used to analyse pressurized water reactor fuel rod behavior during the blowdown phase under large break loss of coolant accident conditions. Two cases have been analysed: without and with initialization from FRAPCON-2 steady state data. (author)

  6. Nuclear fuel cycle facility accident analysis handbook

    The Accident Analysis Handbook (AAH) covers four generic facilities: fuel manufacturing, fuel reprocessing, waste storage/solidification, and spent fuel storage; and six accident types: fire, explosion, tornado, criticality, spill, and equipment failure. These are the accident types considered to make major contributions to the radiological risk from accidents in nuclear fuel cycle facility operations. The AAH will enable the user to calculate source term releases from accident scenarios manually or by computer. A major feature of the AAH is development of accident sample problems to provide input to source term analysis methods and transport computer codes. Sample problems and illustrative examples for different accident types are included in the AAH

  7. PASLOCA: two-dimensional code for loss of primary coolant accident analysis in pool type reactors for use in micro computers

    In order to improve the better performance of the MILOCA code, a 2-dimensional code for Loss of Coolant Accident Analysis in pool type research reactor for use in IBM-PC, an adaptation of the code was made from FORTRAN to PASCAL. This paper presents also the heat transfer model calculation from the fuel elements to the air after the draining of the pool water. As an example, the analysis made for the IAEA 2 MW reference reactor is presented. Differences in computing time between the two versions are also shown. (author)

  8. Upgrade of a fusion accident analysis code and its application to a comparative study of seven fusion reactor designs

    Fusion energy has the potential to be a safe and environmentally favorable energy source. The importance of safety necessitates the existence of a computer code having the capability of assessing off-site impacts resulting from postulated fusion reactor accidents. The FUSCRAC3 computer code has been developed for this purpose. FUSCRAC3 calculates doses resulting from inhalation, groundshine, and cloudshine for 259 isotopes as well as doses resulting from ingestion for 145 isotopes. FUSCRAC3's data base includes the most up-to-date dose conversion factors for all four exposure pathways as well as the most current environmental transfer factors for the ingestion pathway. This work presents a detailed description of the modifications made to the existing fusion reactor accident code, FUSCRAC2, in the development of the more advanced FUSCRAC3 computer code. Also included is a report of the validation procedures. Finally, the improved computer code was applied in two ways: to provide a general data base presenting rem per curie data for each isotope and to assess the doses resulting from possible releases from the reactors evaluated in the ESECOM study. Regarding the latter application, it was found that the general trends established in the original study remained unchanged. However, it was determined that the inclusion of the ingestion pathway substantially affects the overall chronic dose. Isotopes of particular interest due to the ingestion contribution include H-3, Ca-45, Fe-55, and Po-210. 12 refs., 2 figs., 12 tabs

  9. Use and development of coupled computer codes for the analysis of accidents at nuclear power plants. Proceedings of a technical meeting

    Computer codes are widely used in Member States for the analysis of safety at nuclear power plants (NPPs). Coupling of computer codes, a further tool for safety analysis, is especially beneficial to safety analysis. The significantly increased capacity of new computation technology has made it possible to switch to a newer generation of computer codes, which are capable of representing physical phenomena in detail and include a more precise consideration of multidimensional effects. The coupling of advanced, best estimate computer codes is an efficient method of addressing the multidisciplinary nature of reactor accidents with complex interfaces between disciplines. Coupling of computer codes is very advantageous for studies which relate to licensing of new NPPs, safety upgrading programmes for existing plants, periodic safety reviews, renewal of operating licences, use of safety margins for reactor power uprating, better utilization of nuclear fuel and higher operational flexibility, justification for lifetime extensions, development of new emergency operating procedures, analysis of operational events and development of accident management programmes. In this connection, the OECD/NEA Working Group on the Analysis and Management of Accidents (GAMA) recently highlighted the application of coupled computer codes as an area of 'high collective interest'. Coupled computer codes are being developed in many Member States independently or within small groups composed of several technical organizations. These developments revealed that there are many types and methods of code coupling. In this context, it was believed that an exchange of views and experience while addressing these problems at an international meeting could contribute to the more efficient and reliable use of advanced computer codes in nuclear safety applications. The present publication constitutes the report on the Technical Meeting on Progress in the Development and Use of Coupled Codes for Accident

  10. DOZIM - evaluation dose code for nuclear accident

    During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 102 to 2 x 104 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on a

  11. Validation of CONTAIN-LMR code for accident analysis of sodium-cooled fast reactor containments

    CONTAIN-LMR 1 is an analytical tool for the containment performance of sodium cooled fast reactors. In this code, the modelling for the sodium fire is included: the oxygen diffusion model for the sodium pool fire, and the liquid droplet model for the sodium spray fire. CONTAIN-LMR is also able to model the interaction of liquid sodium with concrete structure. It may be applicable to different concrete compositions. Testing and validation of these models will help to qualify the simulation results. Three experiments with sodium performed in the FAUNA facility at FZK have been used for the validation of CONTAIN-LMR. For pool fire tests, calculations have been performed with two models. The first model consists of one gas cell representing the volume of the burn compartment. The volume of the second model is subdivided into 32 coupled gas cells. The agreement between calculations and experimental data is acceptable. The detailed pool fire model shows less deviation from experiments. In the spray fire, the direct heating from the sodium burning in the media is dominant. Therefore, single cell modeling is enough to describe the phenomena. Calculation results have reasonable agreement with experimental data. Limitations of the implemented spray model can cause the overestimation of predicted pressure and temperature in the cell atmosphere. The ability of the CONTAIN-LMR to simulate the sodium pool fire accompanied by sodium-concrete reactions was tested using the experimental study of sodium-concrete interactions for construction concrete as well as for shielding concrete. The model provides a reasonably good representation of chemical processes during sodium-concrete interaction. The comparison of time-temperature profiles of sodium and concrete shows, that the model requires modifications for predictions of the test results. (authors)

  12. Rod ejection accident 3D-dynamic analysis in Trillo NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Trillo NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCSv2.7. In this work, we present the results of the REA analysis at 30% of the rated power at BOC. In the thermalhydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 177 x 32 active nodes, considering 28 different fuel elements with 867 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  13. Rod ejection accident 3D-dynamic analysis in Almaraz NPP with RELAP5/PARCS V2.7 coupled codes

    The Rod Ejection Accident (REA) belongs to the Reactivity Initiated Accidents (RIA) category of accidents, and it is part of the licensing basis accident analyses required for pressurized water reactors (PWR). The REA consist of a rod ejection due to the failure of its driving mechanism. The evolution is driven by a continuous reactivity insertion. In previous works, we have analyzed this transient in Almaraz NPP at different power levels at the beginning of cycle (BOC) and at the end of cycle (EOC) using the coupled code RELAP5-MOD3.3/PARCS v2.7. In this work, we present the results of the REA analysis at hot zero power at BOC with all control rods inserted. In the thermal-hydraulic model used, each fuel assemblies has been modelled as an independent channel, for that, the RELAP5 source code has been modified and recompiled to accept this large number of channels. The neutronic nodal discretization consists of 157 x 24 active nodes, considering 13 different fuel elements with 291 neutronic compositions. The cross-sections sets are obtained from CASMO4-SIMULATE3 using the SIMTAB methodology developed in UPV. The transient departs from an initially critical core, being the withdrawal speed of the control rod a typical bounding value. These analyses will allow knowing more accurately the PWR real plant phenomenology in the RIA most limiting conditions, in this way the conclusions will be realistic. The aim is to improve the understanding of these accidents using advanced methods. (authors)

  14. The Application of Paret/ANL Code for Accident Analysis on Inadvertent Control Rod Withdrawal for RSG GAS Reactor

    The analysis is intended to take a look the condition of safety parameters such as fuel and clad temperature, and minimum safety margin against flow instability (S) in the occurrence of inadvertent control rod withdrawal at nominal power, which is performed by PARET/ANL Code. The accident is initiated when all control rods are simultaneously withdrawn with maximum speed of 0.0564 cm/s which consequently gives maximum reactivity insertion rate σρ/σt = 2.82 x 10-4/s, resulting in the Reactor Protection System (RPS) respond to scram the reactor by dropping the control rods into the core. The primary cooling system is assumed to be in normal operation. It is postulated that the first trip signal from over power is not effective to scram the reactor, but only the second signal from Floating Limit Value eventually causes a scram with 0.5 s delays. During the occurrence of inadvertent control rods withdrawal at 30 MW of initial power, the maximum fuel and clad temperature reach 181.29oC and 137.62oC, respectively and the peak power of 37.11 MW. Meanwhile the minimum value of S reaches 2.62. Therefore, during the occurrence of control rods withdrawal at initial power of 30 MW, the integrity of fuel and clad can be maintained secure since they do not exceed the maximum limit of fuel and clad temperature of 207oC and 145oC, respectively and the minimum value of S is still higher than the design limit of 1.48 for anticipated transient

  15. Post-test analysis of two accident management experiments performed at the BETHSY test facility using the code ATHLET

    In the framework of the external validation of the thermal-hydraulic code ATHLET, which has been developed by the GRS, post test analyses of two experiments were done, which were performed at the french integral test facility BETHSY. During the experiment 5.2 C the complete loss of steam generator feedwater was simulated. The de-pressurization of the primary circuit and high pressure injection is assumed as an emergency measure. During the experiment 9.3 the break of a steam generator U-tube is simulated. The failure of the high pressure injection is assumed. As accident management measures, the depressurization of the steam generator secondary sides and finally of the primary circuit by opening of the pressurizer valve were investigated. The results show, that the code ATHLET is able to describe the complex scenario in good accordance with the experiment. For both tests the safety related statement could be reproduced. (author)

  16. Detailed thermalhydraulic analysis of induced break severe accidents using the massively parallel CFD code TrioU/Priceles

    This paper reports the preliminary studies carried out with the CFD (computational fluid dynamics) code TrioU to study the natural gas circulation that may flow in the primary circuit of a pressurized water reactor during a high-pressure severe accident scenario. Two types of 3-dimensional simulations have been performed on one loop using a LES (large eddy simulations) approach. In the first type of calculations, the gas flow in the hot leg has been investigated with a simplified representation of the reactor vessel and the Steam Generator (SG) tubes. Structured and unstructured meshing have been tested on the full-scale geometry with and without radiative heat transfer modelling between walls and gas. The second type of calculations deals with the gas circulation in the SG. The first results show a good agreement with the available experimental data and provide some confidence in the TrioU code to simulate complex natural flows. (authors)

  17. Metal oxide aerosol dry deposition in laminar pipe flow at high thermal gradients and comparison with SOPHAEROS module of ASTEC reactor accident analysis code

    Highlights: • Experiments to simulate aerosol deposition in pipes in severe accident condition. • Laminar flow conditions under high thermal gradient have been used. • Results of the experiments have been compared with SOPHAEROS module of the ASTEC code. - Abstract: During a severe nuclear reactor accident involving core melt down conditions, the deposition of fission product aerosols inside the reactor coolant system affects the final source term available to the containment and subsequently to the environment. Towards quantifying the aerosol deposition under varying flow conditions and thermal gradients, as may be encountered in the heat transport systems, experiments were performed to investigate the dry deposition behavior of metal oxide aerosols in a 3.6 m long stainless steel piping test assembly. This assembly consisted of divergent and convergent sections, horizontal and vertical sections and right angle bends. Tin oxide aerosols, generated by a plasma torch aerosol generator, were transported into the test assembly using argon carrier gas. Temperature sensors coupled to data loggers were used to record the pipe inner wall and carrier gas temperatures. The experimental deposition results were found to be within 8% of those estimated by the SOPHAEROS module of the accident analysis code ASTEC (Accident Source Term Evaluation Code). Code results for experimental input parameters showed that for sections at higher temperature gradients the dominant deposition mechanism was thermophoresis, while in sections for low thermal gradients, gravitational settling dominated. The micrographs obtained using Transmission Electron Microscopy (TEM) showed that the deposited Tin oxide particles were mostly spherical and bimodal in nature. The X-ray diffraction (XRD) analysis showed that plasma torch generated aerosols exhibit tetragonal SnO and SnO2 phases

  18. Bases of general calculation thermohydrodynamic means (codes) verification and validation methodology for accident analysis at nuclear power plants

    On the basis of previous known approaches' analysis the generalised calculation thermohydrodynamic means verification/validation (V/V) methodology for accident/transition processes' analysis at NPPs is offered in this article. Taking into account formulated requirements and principles the basic V/V procedures, their correlation and order are grounded and considered. The realisation order includes forming calculation means applicability assessment criteria system, analysing mathematical models adequacy to real processes, developing test data bases including a stands adequacy analysis to full-scale conditions, results generalisation methods for final calculation means applicability assessments for specific tasks at specific equipment

  19. Evaluation of confinement capability of radioactive materials under the fire accident in nuclear fuel facility with CELVA-1D (Cell Ventillation Analysis Code-1D)

    To reduce the clogging of smoke on the HEPA filters under the fire accident, some of ventilation systems in the plant are equipped with the pre-filters in front of the HEPA filters for collecting the relatively large smoke particles. Appropriate correspondence such as the exchange of the pre-filter is important for confinement of radioactive materials in the ventilation system under the fire accident. To study smoke generation behavior due to the burnable wastes and clogging properties of the ventilation filters by smoke loading, the verification test has been performed. The cell ventilation system analysis code, CELVA-1D was used for analysis of smoke generation and the rising of pressure drop at both the pre-filter and the HEPA filter. With the change of source term, the breakage of time of the pre-filter was also estimated. (author)

  20. Severe accident analysis using dynamic accident progression event trees

    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

  1. Code strategy for simulating Severe Accident Scenario

    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)

  2. Test Data for USEPR Severe Accident Code Validation

    J. L. Rempe

    2007-05-01

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

  3. Code assessment and modelling for Design Basis Accident analysis of the European Sodium Fast Reactor design. Part II: Optimised core and representative transients analysis

    Lazaro, A., E-mail: aulach@iqn.upv.es [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Schikorr, M. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Mikityuk, K. [PSI, Paul Scherrer Institut, 5232 Villigen (Switzerland); Ammirabile, L. [JRC-IET European Commission, Westerduinweg 3, PO BOX 2, 1755 ZG Petten (Netherlands); Bandini, G. [ENEA, Via Martiri di Monte Sole 4, 40129 Bologna (Italy); Darmet, G.; Schmitt, D. [EDF, 1 Avenue du Général de Gaulle, 92141 Clamart (France); Dufour, Ph.; Tosello, A. [CEA, St. Paul lez Durance, 13108 Cadarache (France); Gallego, E.; Jimenez, G. [UPM, José Gutiérrez Abascal, 2, 28006 Madrid (Spain); Bubelis, E.; Ponomarev, A.; Kruessmann, R.; Struwe, D. [KIT, Institute for Neutron Physics and Reactor Technology, Hermann-von-Helmholtz-Platz 1, 76344 Eggenstein-Leopoldshafen (Germany); Stempniewicz, M. [NRG, Utrechtseweg 310, P.O. Box-9034, 6800 ES Arnhem (Netherlands)

    2014-10-01

    Highlights: • Benchmarked models have been applied for the analysis of DBA transients of the ESFR design. • Two system codes are able to simulate the behavior of the system beyond sodium boiling. • The optimization of the core design and its influence in the transients’ evolution is described. • The analysis has identified peak values and grace times for the protection system design. - Abstract: The new reactor concepts proposed in the Generation IV International Forum require the development and validation of computational tools able to assess their safety performance. In the first part of this paper the models of the ESFR design developed by several organisations in the framework of the CP-ESFR project were presented and their reliability validated via a benchmarking exercise. This second part of the paper includes the application of those tools for the analysis of design basis accident (DBC) scenarios of the reference design. Further, this paper also introduces the main features of the core optimisation process carried out within the project with the objective to enhance the core safety performance through the reduction of the positive coolant density reactivity effect. The influence of this optimised core design on the reactor safety performance during the previously analysed transients is also discussed. The conclusion provides an overview of the work performed by the partners involved in the project towards the development and enhancement of computational tools specifically tailored to the evaluation of the safety performance of the Generation IV innovative nuclear reactor designs.

  4. Investigation of Local Effects Influence on Results of Design Basis Accident Analysis of WWER-440 Reactor Using RELAP5-3D Code

    One of the most important tasks in today's nuclear power plant safety analysis is a simulation of physical processes at nuclear facilities which accounts for 3-dimensional effects in the core and downcomer of reactor. System coupled thermo-hydraulic/neutron-kinetic code RELAP5-3D, which is a modeling tool provided to University of Kyiv by US DOE in a frame of International Nuclear Safety Program, allows simulation of variable in time spatial distribution of neutron flux in a core and also includes special components for 3D modeling of thermo-hydraulics. A model of Rivne NPP Unit 1 with WWER-440/V-213 type reactor has been developed for RELAP5-3D code. A scenario of 'Main steam line break' design basis accident has been calculated using this model. Such a problem can be characterized by intensive overcooling of a primary coolant in affected loop and, taking into account partial mixing of coolant from different primary loops, a non-uniform cooling of reactor core. Obtained results have been compared with the results obtained by model, which has been used at Design Based Accidents analysis, performed at specified unit.(author)

  5. German (GRS) approach to accident analysis (part III). Status of uncertainty evaluations of thermal-hydraulic code results in Germany

    There is an increasing interest in computational reactor safety analysis to replace the conservative evaluation model calculations by best estimate calculations supplemented by a quantitative uncertainty analysis. Sources of uncertainties - code models, initial and boundary conditions, plant state, fuel parameters, scaling, and numerical solution algorithm. Measurements, which are the basis of computer code model, show a scatter around a mean value. For example, data for two-phase pressure drop show a scatter range of about ± 20 - 30%. A range of values should be taken into account for the respective model parameter instead of one discrete value only. The state of knowledge about all uncertain parameters is described by ranges and subjective probability distributions. Stochastic variability due to possible component failures of the reactor plant is not considered in an uncertainty analysis. The single failure criterion is taken into account in a deterministic way. The aim of the uncertainty analysis is at first to identify and quantify all potentially important uncertain parameters. Their propagation through computer code calculations provides subjective probability distributions (and ranges) for the code results. The evaluation of the margin to acceptance criteria, (= technical limit value) e.g. the maximum fuel rod clad temperature, should be based on the upper limit of this distribution for the calculated temperatures. Investigations are underway to transform data measured in experiments and post-test calculations into thermal-hydraulic model parameters with uncertainties. It is effective to concentrate on those uncertainties showing the highest sensitivity measures. The state of knowledge about these uncertain input parameters has to be improved, and suitable experimental as well as analytical information has to be selected. This is a general experience applying different uncertainty methods

  6. Best Estimate Thermal-Hydraulic System Analysis using the MARS Code for the Steam Generator Tube Rupture Accident in the APR1400

    A postulated SGTR (Steam Generator Tube Rupture) accident of the APR1400 was analysed using the best estimate safety analysis code, MARS (Multidimensional Analysis of Reactor Safety). The SGTR accident is one of the design basis accidents, which has a unique feature of the penetration of the barrier between the reactor coolant system (RCS) and the secondary system resulting from the failure of a steam generator U-tube. The SGTR has an importance in safety due to a concern of a containment bypass of radioactive inventory. In the course of the SGTR, the radioactivity leaking from a broken steam generator Utube mixes with the shell-side water in an affected steam generator. Leak flow from ruptured U-tubes can increase a water level and a pressure of the affected steam generator. Following a reactor trip and a turbine trip, the main steam safety valves (MSSVs) can be open to mitigate an increase in the secondary system pressure. Meanwhile, the SGTR can provide a direct flow path from the primary to the secondary system resulting in the release of fission products into the atmosphere. As one of the most limiting SGTR accidents, a leak flow equivalent to a double-ended rupture of five Utubes was analysed in this study. The main objective of this study is not only to provide physical insight into the system response of the APR1400 reactor during a SGTR but also to investigate the effect of reactor trip type of the HSGL (High Steam Generator Level) trip and the LPP (Low Pressurizer Pressure) trip on the thermal-hydraulic system response

  7. Models for describing the behaviour of light water reactors in serious accidents for the programs SCDAP/RELAP5, ATHLET/SA, CATHARE/ICARE, MELCOR etc.. First technical report on BMFT-sponsored research project 1500 831 7: Comparative assessment of different computer codes for severe accident analysis, contribution to the ATHLET/CD code development

    Within the scope of the project BMFT No. 15008317 entitled ''Comparative Assessment of Different Computer Codws for Severe Accident Analysis, Contribution to the ATHLET/SA-Code Development'' the codes ATHLET/SA, CATHARE/ICARE, MELCOR and SCDAP/RELAP5 are investigated. Emphasis is put on a comparison and an assessment of the governing modelling features implemented and operating in the codes under consideration. The codes are evaluated and compared on the base of selected experiments (especially the CORA experimental program of the Karlsruhe Research Center) and relevant severe accident scenarios. The present report is a reference study dealing with the governing models implemented in the severe accident codes SCDAP/RELAP5, ATHLET/SA, CATHARE/ICARE, MELCOR, KESS-III, MAAP and MELPROG/TRAC. Emphaisis is laid on the following models (molstly implemented in form of modules in the respective codes) dealing with: - thermal hydraulics; - heat generation and heat structures; - Radiation heat transfer; - mechanical (rod) behaviour; - core heatup, meltdown and relocation; - chemical reaction; - fission product release and transport; - material properties; - specific components. (orig.)

  8. ATHLET-CD and COCOSYS: the mechanistic computer codes of GRS for simulating severe accidents

    Simulating accident sequences within the framework of safety analyses of nuclear power plants requires the use of deterministic computer codes furnishing the most realistic results (best estimates) in the light of the state of the art. This requirement exists for design basis accidents as well as accidents and events exceeding the design basis. For simulations of reactor behavior and of the source term from the nuclear steam supply system, the ATHLET (Analysis of Thermohydraulics of Leaks and Transients) code has been developed and validated for transients and accidents without major core damage, and the ATHLET-CD (Core Degradation) code has been developed and validated for accidents resulting in major core damage, while the COCOSYS (Containment Code System) code has been developed and validated for the behavior of the containment and the source term for the environment. (orig.)

  9. Verification for flow analysis capability in the model of three-dimensional natural convection with simultaneous spreading, melting and solidification for the debris coolability analysis module in the severe accident analysis code 'SAMPSON', (I)

    The debris coolability analysis module in the severe accident analysis code 'SAMPSON' has been enhanced to predict more mechanistically the safety margin of present reactor pressure vessels in a severe accident. The module calculates debris three-dimensional natural convection with simultaneous spreading, melting and solidification using the 'debris spreading-cooling model' in combination with the temperature distribution of the vessel wall and it evaluates the wall failure. Debris spreading is solved by the free surface calculation method in which the height function is applied. The model makes possible a multiplex heat and mass transfer analysis with flow spearhead and melt front transportation for a single-phase flow analysis code through the resetting of two types of mesh attributions and re-arrangement of the pressure matrix at each time step. The results calculated with the present model are compared with the results from a water spreading experiment. The comparisons verify the model capability for predictions of debris flow in the spreading process. The module provides a good tool for prediction of the reactor safety margin in a severe accident through the three-dimensional natural convection analysis of debris with simultaneous spreading, melting and solidification. (author)

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

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

  11. TRUMP-BD: A computer code for the analysis of nuclear fuel assemblies under severe accident conditions

    TRUMP-BD (Boil Down) is an extension of the TRUMP (Edwards 1972) computer program for the analysis of nuclear fuel assemblies under severe accident conditions. This extension allows prediction of the heat transfer rates, metal-water oxidation rates, fission product release rates, steam generation and consumption rates, and temperature distributions for nuclear fuel assemblies under core uncovery conditions. The heat transfer processes include conduction in solid structures, convection across fluid-solid boundaries, and radiation between interacting surfaces. Metal-water reaction kinetics are modeled with empirical relationships to predict the oxidation rates of steam-exposed Zircaloy and uranium metal. The metal-water oxidation models are parabolic in form with an Arrhenius temperature dependence. Uranium oxidation begins when fuel cladding failure occurs; Zircaloy oxidation occurs continuously at temperatures above 13000 degree F when metal and steam are available. From the metal-water reactions, the hydrogen generation rate, total hydrogen release, and temporal and spatial distribution of oxide formations are computed. Consumption of steam from the oxidation reactions and the effect of hydrogen on the coolant properties is modeled for independent coolant flow channels. Fission product release from exposed uranium metal Zircaloy-clad fuel is modeled using empirical time and temperature relationships that consider the release to be subject to oxidation and volitization/diffusion (''bake-out'') release mechanisms. Release of the volatile species of iodine (I), tellurium (Te), cesium (Ce), ruthenium (Ru), strontium (Sr), zirconium (Zr), cerium (Cr), and barium (Ba) from uranium metal fuel may be modeled

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

    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.

  13. Modeling of pipe break accident in a district heating system using RELAP5 computer code

    Reliability of a district heat supply system is a very important factor. However, accidents are inevitable and they occur due to various reasons, therefore it is necessary to have possibility to evaluate the consequences of possible accidents. This paper demonstrated the capabilities of developed district heating network model (for RELAP5 code) to analyze dynamic processes taking place in the network. A pipe break in a water supply line accident scenario in Kaunas city (Lithuania) heating network is presented in this paper. The results of this case study were used to demonstrate a possibility of the break location identification by pressure decrease propagation in the network. -- Highlights: ► Nuclear reactor accident analysis code RELAP5 was applied for accident analysis in a district heating network. ► Pipe break accident scenario in Kaunas city (Lithuania) district heating network has been analyzed. ► An innovative method of pipe break location identification by pressure-time data is proposed.

  14. Analysis of energy released from core disruptive accident of sodium cooled fast reactor using CDA-ER and VENUS-II codes

    Kang, S. H.; Ha, K. S. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-05-15

    The fast reactor has a unique feature in that rearranged core materials can produce a large increase in reactivity and recriticality. If such a rearrangement of core materials should occur rapidly, there would be a high rate of reactivity increase producing power excursions. The released energy from such an energetic recriticality might challenge the reactor vessel integrity. An analysis of the hypothetical excursions that result in the disassembly of the reactor plays an important role in a liquid metal fast reactor (LMFR) safety analysis. The analysis of such excursions generally consists of three phases (initial or pre-disassembly phase, disassembly phase, energy-work conversion phase). The first step is referred to as the 'accident initiation' or 'pre-disassembly' phase. In this phase, the accident is traced from some initiating event, such as a coolant pump failure or control rod ejection, up to a prompt critical condition where high temperatures and pressures rapidly develop in the core. Such complex processes as fuel pin failure, sodium voiding, and fuel slumping are treated in this phase. Several computer programs are available for this type of calculation, including SAS4A, MELT-II and FREADM. A number of models have been developed for this type of analysis, including the REXCO and SOCOOL-II computer programs. VENUS-II deals with the second phase (disassembly analysis). Most of the models used in the code have been based on the original work of Bethe and Tait. The disassembly motion is calculated using a set of two-dimensional hydrodynamics equations in the VENUS code. The density changes can be explicitly calculated, which in turn allows the use of a more accurate density dependent equation of state. The main functional parts of the computational model can be summarized as follows: Power and energy (point kinetics), Temperature (energy balance), Internal pressure (equation of state), Material displacement (hydrodynamics), Reactivity

  15. Adaption, validation and application of advanced codes with 3-dimensional neutron kinetics for accident analysis calculations - STC with Bulgaria

    In the frame of a project on scientific-technical co-operation funded by BMBF/BMWi, the program code DYN3D and the coupled code ATHLET-DYN3D have been transferred to the Institute for Nuclear Research and Nuclear Energy (INRNE) Sofia. The coupled code represents an implementation of the 3D core model DYN3D developed by FZR into the GRS thermal-hydraulics code system ATHLET. For the purpose of validation of these codes, a measurement data base about a start-up experiment obtained at the unit 6 of Kozloduy NPP (VVER-1000/V-320) has been generated. The results of performed validation calculations were compared with measurement values from the data base. A simplified model for estimation of cross flow mixing between fuel assemblies has been implemented into the program code DYN3D by Bulgarian experts. Using this cross flow model, transient processes with asymmetrical boundary conditions can be analysed more realistic. The validation of the implemented model were performed with help of comparison calculations between modified DYD3D code and thermal-hydraulics code COBRA-4I, and also on the base of the collected measurement data from Kozloduy NPP. (orig.)

  16. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    Bae, Byoung-Uhn, E-mail: bubae@kaeri.re.kr; Kim, Seok; Park, Yu-Sun; Kang, Kyoung-Ho

    2014-08-15

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection.

  17. Integral effect test and code analysis on the cooling performance of the PAFS (passive auxiliary feedwater system) during an FLB (feedwater line break) accident

    Highlights: • This study focuses on the experimental validation of the operational performance of the PAFS (passive auxiliary feedwater system). • A transient simulation of the FLB (feedwater line break) in the integral effect test facility, ATLAS-PAFS, was performed to investigate thermal hydraulic behavior during the PAFS actuation. • The test result confirmed that the APR+ has the capability of coping with the FLB scenario by adopting the PAFS and proper set-points for its operation. • The experimental result was utilized to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. - Abstract: APR+ (Advanced Power Reactor Plus), which is a GEN-III+ nuclear power plant developed in Korea, adopts PAFS (passive auxiliary feedwater system) as an advanced safety feature. The PAFS can completely replace an active auxiliary feedwater system by cooling down the secondary side of steam generators with a natural convection mechanism. This study focuses on experimental and analytical investigation for cooling and operational performance of the PAFS during an FLB (feedwater line break) transient with an integral effect test facility, ATLAS-PAFS. To realistically simulate the FLB accident of the APR+, the three-level scaling methodology was taken into account to design the test facility and determine the test condition. From the test result, the PAFS was actuated to successfully cool down the decay heat of the reactor core by the condensation heat transfer at the PCHX (passive condensation heat exchanger), and thus it could be confirmed that the APR+ has the capability of coping with a FLB scenario by adopting the PAFS and proper set-points for its operation. This integral effect test data were used to evaluate the prediction capability of a thermal hydraulic system analysis code, MARS-KS. The code analysis result proved that it could reasonably predict the FLB transient including the actuation of the PAFS and the natural convection

  18. Coldleg Loss of Coolant Accident (LOCA) Analysis of the Modified Reactor Thermohydraulic Test Facility Using CATHENA Computer Code

    A LOCA analysis at coldleg of the reactor thermal hydraulic test facility using CATHENA computer code has been completely conducted. The analysis was performed by modeling the reactor thermal hydraulic test loop into generic models of the CATHENA such as PUMP, VALVE, VOLUME, ACCUMULATOR, TANK, RESERVOIR, DISCHARGE, GENERALIZED TANK, and GENHTP. The primary system was simulated at power of 1 MWatt with pressure and mass flow at 15.5 MPa and 9.4 kg/s respectively. At secondary side, feedwater flowed at 15.0 kg/s with temperature of 27 oC and pressure of 0.8 MPa. The calculation showed that during steady state the inlet and outlet temperature of test section were 121 oC and 146 oC. After calculating steady state condition, the calculation was followed by transient calculation. The transient was triggered by pipe break at coldleg with diameter of the break was 2 mm. Due to this break, the pressure decreased dramatically. When the pressure reached 4.2 MPa, the accumulator started supplying water into the system. A moment later, the pump was also tripped because of the continuing pressure drop that reached 4.0 MPa. As a consequence the coolant flow was also dropped At the coolant 40 % of normal flow, the power of heated rods then shut down. The result of calculation showed that during the transient, the maximum coolant temperature was 159 oC and the maximum temperature of heated rods was 223 oC. Based on these results, it can be concluded that during the transient, the heated rods were not in danger. (author)

  19. Accident analysis in research reactors

    With the sustained development in computer technology, the possibilities of code capabilities have been enlarged substantially. Consequently, advanced safety evaluations and design optimizations that were not possible few years ago can now be performed. The challenge today is to revisit the safety features of the existing nuclear plants and particularly research reactors in order to verify that the safety requirements are still met and - when necessary - to introduce some amendments not only to meet the new requirements but also to introduce new equipment from recent development of new technologies. The purpose of the present paper is to provide an overview of the accident analysis technology applied to the research reactor, with emphasis given to the capabilities of computational tools. (author)

  20. Analysis of the TMI-2 accident using ATHLET-CD

    One analyzed the simulation of the TMI-2 NPP accident making use of the ATHLET-CD code. One describes the accident sequence, the code structure and performs the comparative analysis of the calculated and the measured data. Simulation of thermohydraulic characteristics was a special success. Application of the codes promotes the NPP optimization, the reactor safety improvement and the risk reduction. The ATHLET-CD system ( the thermohydraulic analysis of leaks and transient processes at the reactor core disruption) will allow to evaluate the adequacy of the models included in the available codes to calculate severe accidents

  1. Experimental Analysis with RANNS Code on Boiling Heat Transfer from Fuel Rod Surface to Coolant Water Under Reactivity-Initiated Accident Conditions

    In order to promote a better understanding of the temperature evolution of fuel rod under reactivity-initiated accident (RIA) conditions, we have investigated the effects of coolant subcooling, flow velocity, pressure, and cladding pre-irradiation on the heat transfer from fuel rod surface to coolant water during RIA boiling transient. The study was based on a computational analysis, with the RANNS code, on the transient data from RIA-simulating experiments in the nuclear safety research reactor (NSRR); boiling heat transfer coefficients were estimated by inverse-heat-conduction calculations using the histories of measured cladding temperature and estimated heat generation in pellets, and the effects of coolant condition were analyzed by a two-phase laminar boundary layer model for stable film boiling. The experimental data used in this study cover coolant conditions with subcoolings of ~10–80 K, flow velocities of 0 to ~3 m/s, pressures of 0.1 to ~16 MPa, and fuel burnups of 0–69 GWd/tU. The analysis showed that the film boiling heat transfer coefficients during RIA boiling transient increase with coolant subcooling, flow velocity, and pressure as predicted by the model for stable film boiling. The estimated boiling heat transfer coefficients were significantly larger than those predicted by semi-empirical correlations for stable film boiling: about 1.5 times larger for stagnant water condition and 2–8 times larger for forced flow condition, respectively. The analysis also suggested that the heat transfers during both transition and film boiling phases are strongly enhanced by pre-irradiation of the cladding. The irradiation effect was clearly seen at large subcooling of ~80 K and atmospheric coolant pressure, and was rather moderate at small subcooling of ~10 K and coolant pressure of ~7 MPa. These behaviors of boiling heat transfer are incorporated into the RANNS code mainly as modified empirical correlations for boiling heat transfer coefficient. (author)

  2. Validation of computer code THYNAC for analysis of loss of coolant accident in pressurised heavy water reactor

    The computer code THYNAC has been validated against the available experimental results of blowdown from RD-4 loop, a Canadian test facility designed to simulate LOCA in PHWR. In this paper available experimental results are compared with predictions made with THYNAC. In general, predictions show consistent trends on pressure transient during LOCA and conservative trends with respect to fuel sheath peak temperatures. (author). 4 refs., 5 figs., 2 tab s

  3. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident

  4. SCDAP/RELAP5/MOD 3.1 code manual: MATPRO, A library of materials properties for Light-Water-Reactor accident analysis. Volume 4

    Hagrman, D.T. [ed.; Allison, C.M.; Berna, G.A. [Lockheed Idaho Technologies Co., Idaho Falls, ID (United States)] [and others

    1995-06-01

    The SCDAP/RELAP5 code has been developed for best estimate transient simulation of light -- water-reactor coolant systems during a severe accident. The code models the coupled behavior of the reactor coolant system, the core, fission products released during a severe accident transient as well as large and small break loss of coolant accidents, operational transients such as anticipated transient without SCRAM, loss of offsite power, loss of feedwater, and loss of flow. A generic modeling approach is used that permits as much of a particular system to be modeled as necessary. Control system and secondary system components are included to permit modeling of plant controls, turbines, condensers, and secondary feedwater conditioning systems. This volume, Volume IV, describes the material properties correlations and computer subroutines (MATPRO) used by SCDAP/RELAP5. formulation of the materials properties are generally semi-empirical in nature. The materials property subroutines contained in this document are for uranium, uranium dioxide, mixed uranium-plutonium dioxide fuel, zircaloy cladding, zirconium dioxide, stainless steel, stainless steel oxide, silver-indium-cadmium alloy, cadmium, boron carbide, Inconel 718, zirconium-uranium-oxygen melts, fill gas mixtures, carbon steel, and tungsten. This document also contains descriptions of the reaction and solution rate models needed to analyze a reactor accident.

  5. Validation of the thermal hydraulic computer code S-RELAP5 for performing loss-of-coolant accident analysis (LOCA) in Pressurized Water Reactors (PWRs)

    Siemens Power Corporation (SPC) has developed S-RELAP5, a RELAP5/MOD2 based thermal hydraulic system code with main modifications and improvements relative to RELAP5/MOD2 concerning Multi-Dimensional Capability, Energy Equations, Numerical Solution of Hydrodynamic, Constitutive Models, Heat Transfer Models, Chocked Flow, and Counter-Current Flow Limiting. S-RELAP5 was exercised over a range of integral and separate effects tests in order to demonstrate that the code could predict the important phenomena associated with PWR LBLOCA. A methodology for calculation of statistical uncertainties has been developed and applied to analyses of hypothetical large break loss-of-coolant accidents (LBLOCA). To extend the application capability of S-RELAP5 to small break loss-of-coolant accidents problems (SBLOCA) an investigation program for appropriate experiments was launched and largely carried out. (author)

  6. Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code

    Abbasi, Molood; Rahgoshay, Mohhamad [Islamic Azad Univ., Teheran (Iran, Islamic Republic of). Dept. of Nuclear Engineering, Science and Research Branch

    2013-11-15

    In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally Guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident. (orig.)

  7. Modeling of DECL accident in the reactor containment by the CONTAIN 2.0 code

    In this paper, a specific type of the Loss of Coolant Accident (LOCA), the DECL (Double Ended Cold Leg) break, that means totally Guillotine type of break in the cold leg pipe, has been modeled. The accident is simulated with the CONTAIN 2.0 code. In the event of a LOCA accident, coolant mass and energy are released to the containment through the break. This causes an increase of pressure and temperature in the containment. The modeling is performed in the VVER-1000 reactor containment. The analysis includes average pressure in the containment and temperature distribution in sample cells in the long-time. The results are compared with the existing reports on studies that used the ANGAR code. Results show that the CONTAIN 2.0 code is an adaptable tool for the analysis of nuclear events such as DECL accident. (orig.)

  8. Accident Tolerant Fuel Analysis

    Curtis Smith; Heather Chichester; Jesse Johns; Melissa Teague; Michael Tonks; Robert Youngblood

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced “RISMC toolkit” that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional “accident-tolerant” (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant decision makers should propose and

  9. Accident tolerant fuel analysis

    Smith, Curtis [Idaho National Laboratory; Chichester, Heather [Idaho National Laboratory; Johns, Jesse [Texas A& M University; Teague, Melissa [Idaho National Laboratory; Tonks, Michael Idaho National Laboratory; Youngblood, Robert [Idaho National Laboratory

    2014-09-01

    Safety is central to the design, licensing, operation, and economics of Nuclear Power Plants (NPPs). Consequently, the ability to better characterize and quantify safety margin holds the key to improved decision making about light water reactor design, operation, and plant life extension. A systematic approach to characterization of safety margins and the subsequent margins management options represents a vital input to the licensee and regulatory analysis and decision making that will be involved. The purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development (R&D) is to support plant decisions for risk-informed margins management by improving economics and reliability, and sustaining safety, of current NPPs. Goals of the RISMC Pathway are twofold: (1) Develop and demonstrate a risk-assessment method coupled to safety margin quantification that can be used by NPP decision makers as part of their margin recovery strategies. (2) Create an advanced ''RISMC toolkit'' that enables more accurate representation of NPP safety margin. In order to carry out the R&D needed for the Pathway, the Idaho National Laboratory is performing a series of case studies that will explore methods- and tools-development issues, in addition to being of current interest in their own right. One such study is a comparative analysis of safety margins of plants using different fuel cladding types: specifically, a comparison between current-technology Zircaloy cladding and a notional ''accident-tolerant'' (e.g., SiC-based) cladding. The present report begins the process of applying capabilities that are still under development to the problem of assessing new fuel designs. The approach and lessons learned from this case study will be included in future Technical Basis Guides produced by the RISMC Pathway. These guides will be the mechanism for developing the specifications for RISMC tools and for defining how plant

  10. Thermal-hydraulic analysis best-estimate of an accident in the containment a PWR-W reactor with GOTHIC code using a 3D model detailed; Analisis termo-hidraulico best-estimate de un accidente en contencion de un reactor PWR-W con el codigo GOTHIC mediante un modelo 3D detallado

    Bocanegra, R.; Jimenez, G.

    2013-07-01

    The objective of this project will be a model of containment PWR-W with the GOTHIC code that allows analyzing the behavior detailed after a design basis accident or a severe accident. Unlike the models normally used in codes of this type, the analysis will take place using a three-dimensional model of the containment, being this much more accurate.

  11. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    Jin, Hyung Gon; Lee, Dong Won; Lee, Eo Hwak; Yoon, Jae Sung; Kim, Suk Kwon [KAERI, Daejeon (Korea, Republic of); Merrill, Brad J. [Idaho National Laboratory, Atomic (United States); Ahn, Mu-Young; Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of)

    2015-05-15

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'.

  12. Status of the GAMMA-FR code validation - TES pipe rupture accident of HCCR TBS

    GAMMA-FR code to code validation is conducted and it shows reasonable agreement, however, near wall effect on the effective thermal conductivity needs to be investigated for better results. The GAMMA-FR code was scheduled for validation during the next two years under UCLA-NFRI collaboration. Through this research, GAMMA-FR will be validated with representative fusion experiments and reference accident cases. The GAMMA-FR (Gas Multicomponent Mixture Transient Analysis for Fusion Reactors) code is an in-house system analysis code to predict the thermal hydraulic and chemical reaction phenomena expected to occur during the thermo-fluid transients in a nuclear fusion system. A safety analysis of the Korea TBS (Test Blanket System) for ITER (International Thermonuclear Experimental Reactor) is underway using this code. This paper describes validation strategy of GAMMA-FR and current status of the validation study with respect to 'TES pipe rupture accident of ITER TBM'

  13. Analysis of an accident type sbloca in reactor contention AP1000 with 8.0 Gothic code; Analisis de un accidente tipo Sbloca en la contencion del reactor AP1000 con el codigo Gothic 8.0

    Goni, Z.; Jimenez Varas, G.; Fernandez, K.; Queral, C.; Montero, J.

    2016-08-01

    The analysis is based on the simulation of a Small Break Loss-of-Coolant-Accident in the AP1000 nuclear reactor using a Gothic 8.0 tri dimensional model created in the Science and Technology Group of Nuclear Fision Advanced Systems of the UPM. The SBLOCA has been simulated with TRACE 5.0 code. The main purpose of this work is the study of the thermo-hydraulic behaviour of the AP1000 containment during a SBLOCA. The transients simulated reveal close results to the realistic behaviour in case of an accident with similar characteristics. The pressure and temperature evolution enables the identification of the accident phases from the RCS point of view. Compared to the licensing calculations included in the AP1000 Safety Analysis, it has been proved that the average pressure and temperature evolution is similar, yet lower than the licensing calculations. However, the temperature and inventory distribution are significantly heterogeneous. (Author)

  14. The development of severe accident analysis technology

    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)

  15. The development of severe accident analysis technology

    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).

  16. Codes, methods and approaches for accident analyses of the core and fuel behaviour

    Thermohydraulic and fuel behaviour computer codes developed for WWER reactors by the Nuclear Power Plants Research Institute, Trnava (SK), are described. The features of presently used codes PIN, DEFOS-1A, DEFOS-2A, SICHTA, FEMBUL, CALOPEA and DYN3D/M3, their utilization areas, interconnections and the safety analyses procedures are briefly described. General approach in safety simulation and evaluation is given. The interconnections between the proposed criteria - anticipated transients, postulated accidents and cladding failure - are shown. The acceptance criteria of IAEA are checked by the analyses of the transients using the corresponding codes. For most accident analyses, the transient simulation by means of the codes for system transient analysis (RECAP, DYNAMIKA etc.) is sufficient to provide evaluation of the criteria needed. For some transients more detailed analysis is necessary using DYN3D and SICHTA codes (e.g., reactivity initiated accidents). Parameters defining fuel behaviour are determined having in mind that for most of the typical WWER accidents no or very limited damage of fuel assemblies occurred. It allows, on one hand, the use of conservative criteria, and, on the other, to use approach of bounding accidents for proving some criteria like calculated doses below limits, local clad oxidation not exceeding 17% and hydrogen generation below limit. It limits in the current conditions the necessary use of PIN and DEFOS codes to not very large number of analyses. 1 tab., 8 refs

  17. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    Ball, S.J.; Cleveland, J.C.; Sanders, J.P.

    1978-04-04

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes.

  18. Evaluation of the General Atomic codes TAP and RECA for HTGR accident analyses

    The General Atomic codes TAP (Transient Analysis Program) and RECA (Reactor Emergency Cooling Analysis) are evaluated with respect to their capability for predicting the dynamic behavior of high-temperature gas-cooled reactors (HTGRs) for postulated accident conditions. Several apparent modeling problems are noted, and the susceptibility of the codes to misuse and input errors is discussed. A critique of code verification plans is also included. The several cases where direct comparisons could be made between TAP/RECA calculations and those based on other independently developed codes indicated generally good agreement, thus contributing to the credibility of the codes

  19. Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty; Qualificacao e aplicacao de codigo de acidentes de reatores nucleares com capacidade interna de avaliacao de incerteza

    Borges, Ronaldo Celem

    2001-10-15

    This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

  20. Incorporation of advanced accident analysis methodology into safety analysis reports

    The IAEA Safety Guide on Safety Assessment and Verification defines that the aim of the safety analysis should be by means of appropriate analytical tools to establish and confirm the design basis for the items important to safety, and to ensure that the overall plant design is capable of meeting the prescribed and acceptable limits for radiation doses and releases for each plant condition category. Practical guidance on how to perform accident analyses of nuclear power plants (NPPs) is provided by the IAEA Safety Report on Accident Analysis for Nuclear Power Plants. The safety analyses are performed both in the form of deterministic and probabilistic analyses for NPPs. It is customary to refer to deterministic safety analyses as accident analyses. This report discusses the aspects of using the advanced accident analysis methods to carry out accident analyses in order to introduce them into the Safety Analysis Reports (SARs). In relation to the SAR, purposes of deterministic safety analysis can be further specified as (1) to demonstrate compliance with specific regulatory acceptance criteria; (2) to complement other analyses and evaluations in defining a complete set of design and operating requirements; (3) to identify and quantify limiting safety system set points and limiting conditions for operation to be used in the NPP limits and conditions; (4) to justify appropriateness of the technical solutions employed in the fulfillment of predetermined safety requirements. The essential parts of accident analyses are performed by applying sophisticated computer code packages, which have been specifically developed for this purpose. These code packages include mainly thermal-hydraulic system codes and reactor dynamics codes meant for the transient and accident analyses. There are also specific codes such as those for the containment thermal-hydraulics, for the radiological consequences and for severe accident analyses. In some cases, codes of a more general nature such

  1. Qualification of the nuclear reactor core model DYN3D coupled to the thermohydraulic system code ATHLET, applied as an advanced tool for accident analysis of VVER-type reactors. Final report

    The nuclear reactor core model DYN3D with 3D neutron kinetics has been coupled to the thermohydraulic system code ATHLET. In the report, activities on qualification of the coupled code complex ATHLET-DYN3D as a validated tool for the accident analysis of russian VVER type reactors are described. That includes: - Contributions to the validation of the single codes ATHLET and DYN3D by the analysis of experiments on natural circulation behaviour in thermohydraulic test facilities and solution of benchmark tasks on reactivity initiated transients, - the acquisition and evaluation of measurement data on transients in nuclear power plants, the validation of ATHLET-DYN3D by calculating an accident with delayed scram and a pump trip in VVER plants, - the complementary improvement of the code DYN3D by extension of the neutron physical data base, implementation of an improved coolant mixing model, consideration of decay heat release and xenon transients, - the analysis of steam leak scenarios for VVER-440 type reactors with failure of different safety systems, investigation of different model options. The analyses showed, that with realistic coolant mixing modelling in the downcomer and the lower plenum, recriticality of the scramed reactor due to overcooling can be reached. The application of the code complex ATHLET-DYN3D in Czech Republic, Bulgaria and the Ukraine has been started. Future work comprises the verification of ATHLET-DYN3D with a DYN3D version for the square fuel element geometry of western PWR. (orig.)

  2. Radioactive materials transport accident analysis

    Over the last 25 years, one of the major issues raised regarding radioactive material transportation has been the risk of severe accidents. While numerous studies have shown that traffic fatalities dominate the risk, modeling the risk of severe accidents has remained one of the most difficult analysis problems. This paper will show how models that were developed for nuclear spent fuel transport accident analysis can be adopted to obtain estimates of release fractions for other types of radioactive material such as vitrified highlevel radioactive waste. The paper will also show how some experimental results from fire experiments involving low level waste packaging can be used in modeling transport accident analysis with this waste form. The results of the analysis enable an analyst to clearly show the differences in the release fractions as a function of accident severity. The paper will also show that by placing the data in a database such as ACCESS trademark, it is possible to obtain risk measures for transporting the waste forms along proposed routes from the generator site to potential final disposal sites

  3. Analysis of severe accidents in pressurized heavy water reactors

    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

  4. Code assessment in context of severe accident phenomenology

    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.)

  5. Coupled code calculation of rod withdrawal at power accident

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced

  6. Coupled code calculation of rod withdrawal at power accident

    Grgić, Davor, E-mail: davor.grgic@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Benčik, Vesna, E-mail: vesna.bencik@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia); Šadek, Siniša, E-mail: sinisa.sadek@fer.hr [Faculty of Electrical Engineering and Computing, University of Zagreb, Unska 3, 10000 Zagreb (Croatia)

    2013-08-15

    Highlights: ► Sensitivity calculations (withdrawal speed, initial power, secondary side influence) were performed for the rod withdrawal at power accident in PWR. ► Best estimate coupled RELAP5-PARCS code calculation was done, using COBRA code to model the core thermal-hydraulics. ► Specific modelling features included reactor vessel split model, explicit model of the RTD bypass and the overtemperature ΔT setpoint function. ► Average whole core values and the local hot spots were predicted. ► Local fuel centerline temperature and local DNBR were calculated using a COBRA-like model. ► Influence of the burnup on the fuel centerline temperature was studied. -- Abstract: The rod withdrawal at power (RWAP) accident is analyzed for NPP Krško as part of activity related to possible resistance temperature detectors (RTDs) bypass removal. The RWAP accident can be departure from nucleate boiling (DNB) or overpower limiting accident depending on initial power level and rate and amount of reactivity addition. In this paper we have analyzed the response of the plant in current configuration to RWAP for different withdrawal speeds and different initial power levels. By demonstrating adequacy of current protection system we can, in the next step, quantify the influence of change in narrow range coolant temperature measurement to available safety margins. The overtemperature ΔT setpoint and its relation to local DNBR values are in center of attention. The coupled RELAP5–PARCS code was used as the calculation tool with the provision to extend the calculation to local pin-by-pin COBRA subchannel calculation for selected state points derived from main coupled code results. In the first part of the calculation methodology, point kinetics calculation is performed using standalone RELAP5 to reproduce USAR results, and in the second part, more demanding coupled code calculation is introduced.

  7. Quest for the real-time for the safety analysis code Cathare 2 used in the post-accident simulator Sipa

    The aim of the SCAR project is to use the CATHARE French thermal-hydraulic accident code in the SIPA simulator (Post-Accident Simulator) and extend SIPA to reactor cold shutdown states. The quest for real-time has been one of the key themes of the project since it began in 1997. The required CPU time depends on the computing power and on the ability of CATHARE to converge as fast as possible on the solution. Three main tasks have been scheduled to contain the lag between the simulation and the real-time: -1) Parallelism in CATHARE has been developed with shared-memory model (using OPEN MP). Standardized and adapted to the numerical method and the structure of CATHARE, it has enabled parallel tasks in 95% of the code with efficient parallel loops on the elements, and an optimized but limited parallelism in the solver. Validation has been carried out all along the task, ensuring the binary identity of results for 10 representative accident transients, whatever the number of processors used on each computer of the SCAR project. -2) Convergence has been improved for 20 CATHARE transients, ranging from the 100% full power state to cold-shutdown for maintenance state. A method based on the definition of maximum lag criteria in function of an estimated power of computers has been developed, revealing coding errors and leading to numerical improvements without any regression of physical law validation. A second phase has started in 2003 on another series of 25 transients within the simulator. -3) A techno-watch policy (using benchmarking) has allowed to keep up to date with progress in computer power throughout the duration of the project. It has consisted in comparing the performance of computers for 12 standard CATHARE input decks using an elementary time relevant of the computing machines for a given modeling of plant series. Furthermore, development validation and performance assessment tools have been developed at the same time. As a result of these three tasks

  8. CONTAIN-LMR程序中池式钠火事故分析计算模型的验证%Verification of sodium pool fire accident analysis model in CONTAIN-LMR code

    李世锐; 任丽霞; 胡文军; 乔鹏瑞

    2016-01-01

    CONTAIN-LMR是针对以液态钠为冷却剂的反应堆而开发的安全壳事故一体化分析程序。我国目前的CONTAIN-LMR程序版本为2000年左右从法国引进,还未进行过面向工程设计的系统性地程序开发和验证。本文主要针对 CONTAIN-LMR 程序中模拟池式钠火事故的分析模型进行详细分析,并采用国际上的池式钠火实验进行验证,实验验证结果表明 CONTAIN-LMR 程序可以较准确地模拟池式钠火事故造成的钠工艺间内的温度、压力升高及放射性钠气溶胶行为。本文的研究结果初步表明CONTAIN-LMR程序可用于钠冷快堆的钠火事故分析。%CONTAIN-LMR is an integrated code which aims at sodium cooled fast reactor containment accident analysis. The current version of the CONTAIN-LMR code in China was imported from France around 2000,program development and verification of engineering level design has not undertaken systematically. This paper makes a detailed analysis for the models of sodium pool fire accident simulation in CONTAIN-LMR code,and uses international sodium pool fire experiments for verification,the result shows that the CONTAIN-LMR code can simulate the temperature,pressure rising and radioactive sodium aerosol behavior in containment caused by sodium pool fire accidents. The studies in this paper indicated that the CONTAIN-LMR code can be used for the analysis of sodium fire accidents in sodium cooled fast reactor.

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

    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)

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

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

    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

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

    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.

  13. Accident analysis and DOE criteria

    In analyzing the radiological consequences of major accidents at DOE facilities one finds that many facilities fall so far below the limits of DOE Order 6430 that compliance is easily demonstrated by simple analysis. For those cases where the amount of radioactive material and the dispersive energy available are enough for accident consequences to approach the limits, the models and assumptions used become critical. In some cases the models themselves are the difference between meeting the criteria or not meeting them. Further, in one case, we found that not only did the selection of models determine compliance but the selection of applicable criteria from different chapters of Order 6430 also made the difference. DOE has recognized the problem of different criteria in different chapters applying to one facility, and has proceeded to make changes for the sake of consistency. We have proposed to outline the specific steps needed in an accident analysis and suggest appropriate models, parameters, and assumptions. As a result we feed DOE siting and design criteria will be more fairly and consistently applied

  14. Adaptation of the ASTEC code system to accident scenarios in fusion installations

    Highlights: ► IRSN has a first version of ASTEC able to model an accident in ITER. ► Models are developed to make possible water/air ingress simulations in the vessel. ► Some thermal-hydraulic calculations in agreement with MELCOR are discussed. -- Abstract: ASTEC is a code system aiming to compute severe accident scenarios and their consequences in nuclear fission Pressurized Water Reactors (PWRs). Its capabilities have been recently extended to address the main accident sequences which may occur in the fusion installations, in particular in ITER. The purpose of this paper is to present a synthesis of the work that has been performed on ASTEC as part of its adaptation to fusion ITER facility, in particular concerning the development of some specific models (dust behavior, jet impaction and wall oxidation), the state of validation of the code and some first calculations for accident transients considered in the basis design. Comparisons with the MELCOR code, selected by ITER for their own safety analysis are provided and show a good agreement between both codes

  15. Probabilistic accident sequence recovery analysis

    Recovery analysis is a method that considers alternative strategies for preventing accidents in nuclear power plants during probabilistic risk assessment (PRA). Consideration of possible recovery actions in PRAs has been controversial, and there seems to be a widely held belief among PRA practitioners, utility staff, plant operators, and regulators that the results of recovery analysis should be skeptically viewed. This paper provides a framework for discussing recovery strategies, thus lending credibility to the process and enhancing regulatory acceptance of PRA results and conclusions. (author)

  16. Development of Database for Accident Analysis in Indian Mines

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

    Mining is a hazardous industry and high accident rates associated with underground mining is a cause of deep concern. Technological developments notwithstanding, rate of fatal accidents and reportable incidents have not shown corresponding levels of decline. This paper argues that adoption of appropriate safety standards by both mine management and the government may result in appreciable reduction in accident frequency. This can be achieved by using the technology in improving the working conditions, sensitising workers and managers about causes and prevention of accidents. Inputs required for a detailed analysis of an accident include information on location, time, type, cost of accident, victim, nature of injury, personal and environmental factors etc. Such information can be generated from data available in the standard coded accident report form. This paper presents a web based application for accident analysis in Indian mines during 2001-2013. An accident database (SafeStat) prototype based on Intranet of the TCP/IP agreement, as developed by the authors, is also discussed.

  17. Visual and intelligent transients and accidents analyzer based on thermal-hydraulic system code

    Full text of publication follows: Many thermal-hydraulic system codes were developed in the past twenty years, such as RELAP5, RETRAN, ATHLET, etc. Because of their general and advanced features in thermal-hydraulic computation, they are widely used in the world to analyze transients and accidents. But there are following disadvantages for most of these original thermal-hydraulic system codes. Firstly, because models are built through input decks, so the input files are complex and non-figurative, and the style of input decks is various for different users and models. Secondly, results are shown in off-line data file form. It is not convenient for analysts who may pay more attention to dynamic parameters trend and changing. Thirdly, there are few interfaces with other program in these original thermal-hydraulic system codes. This restricts the codes expanding. The subject of this paper is to develop a powerful analyzer based on these thermal-hydraulic system codes to analyze transients and accidents more simply, accurately and fleetly. Firstly, modeling is visual and intelligent. Users build the thermalhydraulic system model using component objects according to their needs, and it is not necessary for them to face bald input decks. The style of input decks created automatically by the analyzer is unified and can be accepted easily by other people. Secondly, parameters concerned by analyst can be dynamically communicated to show or even change. Thirdly, the analyzer provide interface with other programs for the thermal-hydraulic system code. Thus parallel computation between thermal-hydraulic system code and other programs become possible. In conclusion, through visual and intelligent method, the analyzer based on general and advanced thermal-hydraulic system codes can be used to analysis transients and accidents more effectively. The main purpose of this paper is to present developmental activities, assessment and application results of the visual and intelligent

  18. Analysis of tritium mission FMEF/FAA fuel handling accidents

    The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix

  19. Analysis of tritium mission FMEF/FAA fuel handling accidents

    Van Keuren, J.C.

    1997-11-18

    The Fuels Material Examination Facility/Fuel Assembly Area is proposed to be used for fabrication of mixed oxide fuel to support the Fast Flux Test Facility (FFTF) tritium/medical isotope mission. The plutonium isotope mix for the new mission is different than that analyzed in the FMEF safety analysis report. A reanalysis was performed of three representative accidents for the revised plutonium mix to determine the impact on the safety analysis. Current versions computer codes and meterology data files were used for the analysis. The revised accidents were a criticality, an explosion in a glovebox, and a tornado. The analysis concluded that risk guidelines were met with the revised plutonium mix.

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

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

  1. Computer code for the analyses of reactivity initiated accident of heavy water moderated and cooled research reactor 'EUREKA-2D'

    Codes, such as EUREKA and EUREKA-2 have been developed to analyze the reactivity initiated accident for light water reactor. These codes could not be applied directly for the analyses of heavy water moderated and cooled research reactor which are different from light water reactor not only on operation condition but also on reactor kinetic constants. EUREKA-2D which is modified EUREKA-2 is a code for the analyses of reactivity initiated accident of heavy water research reactors. Following items are modified: 1) reactor kinetic constants. 2) thermodynamic properties of coolant. 3) heat transfer equations. The feature of EUREKA-2D and an example of analysis are described in this report. (author)

  2. Analysis of local subassembly accident in KALIMER

    Kwon, Young Min; Jeong, Kwan Seong; Hahn, Do Hee

    2000-10-01

    Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage.

  3. Analysis of local subassembly accident in KALIMER

    Subassembly Accidents (S-A) in the Liquid Metal Reactor (LMR) may cause extensive clad and fuel melting and are thus regarded as a potential whole core accident initiator. The possibility of S-A occurrence must be very low frequency by the design features, and reactor must have specific instrumentation to interrupt the S-A sequences by causing a reactor shutdown. The evaluation of the relevant initiators, the event sequences which follow them, and their detection are the essence of the safety issue. Particularly, the phenomena of flow blockage caused by foreign materials and/or the debris from the failed fuel pin have been researched world-widely. The foreign strategies for dealing with the S-A and the associated safety issues with experimental and theoretical R and D results are reviewed. This report aims at obtaining information to reasonably evaluate the thermal-hydraulic effect of S-A for a wire-wrapped LMR fuel pin bundle. The mechanism of blockage formation and growth within a pin bundle and at the subassembly entrance is reviewed in the phenomenological aspect. Knowledge about the recent LMR subassembly design and operation procedure to prevent flow blockage will be reflected for KALIMER design later. The blockage analysis method including computer codes and related analytical models are reviewed. Especially SABRE4 code is discussed in detail. Preliminary analyses of flow blockage within a 271-pin driver subassembly have been performed using the SABRE4 computer code. As a result no sodium boiling occurred for the central 24-subchannel blockage as well as 6-subchannel blockage

  4. Simulation of rod ejection accident in a WWER-1000 Nuclear Reactor by using PARCS code

    Highlights: • REA in WWER-1000 Nuclear Reactor was simulated. • PARCS v2.7 and WIMSD-5B codes were used. • PARCS was validated for steady-state and transient processes. • Temperature reactivity coefficient was calculated. • TH block of PARCS v2.7 code was used. - Abstract: The rod ejection accident is defined as the postulated rupture of a control rod drive mechanism housing that results in the complete ejection of a rod cluster control assembly from the reactor core. The consequences of the mechanical failure are a rapid positive reactivity insertion and an increase in the local power peaking with high local energy deposition in the fuel assembly, accompanied by an initial pressure increase in the reactor cooling system. In this study, the REA has been simulated in a WWER-1000 reactor by using WIMSD-5B and PARCS v2.7 codes. First, macroscopic cross-sections have been calculated for various types of fuel assemblies using WIMSD-5B. Results have been fed as input to PARCS v2.7 code. Steady-state, transient and specially thermal–hydraulic feedback blocks of PARCS code have been handled in this simulation. Finally, results have been compared with Final Safety Analysis Report of WWER-1000 reactor. The results show a great similarity and confirm the ability of PARCS code in simulation of transient accidents

  5. A methodology for radiological accidents analysis in industrial gamma radiography

    A critical review of 34 published severe radiological accidents in industrial gamma radiography, that happened in 15 countries, from 1960 to 1988, was performed. The most frequent causes, consequences and dose estimation methods were analysed, aiming to stablish better procedures of radiation safety and accidents analysis. The objective of this work is to elaborate a radiological accidents analysis methodology in industrial gamma radiography. The suggested methodology will enable professionals to determine the true causes of the event and to estimate the dose with a good certainty. The technical analytical tree, recommended by International Atomic Energy Agency to perform radiation protection and nuclear safety programs, was adopted in the elaboration of the suggested methodology. The viability of the use of the Electron Gamma Shower 4 Computer Code System to calculate the absorbed dose in radiological accidents in industrial gamma radiography, mainly at sup(192)Ir radioactive source handling situations was also studied. (author)

  6. SHETEMP: a computer code for calculation of fuel temperature behavior under reactivity initiated accidents

    A fast running computer code SHETEMP has been developed for analysis of reactivity initiated accidents under constant core cooling conditions such as coolant temperature and heat transfer coefficient on fuel rods. This code can predict core power and fuel temperature behaviours. A control rod movement can be taken into account in power control system. The objective of the code is to provide fast running capability with easy handling of the code required for audit and design calculations where a large number of calculations are performed for parameter surveys during short time period. The fast running capability of the code was realized by neglection of fluid flow calculation. The computer code SHETEMP was made up by extracting and conglomerating routines for reactor kinetics and heat conduction in the transient reactor thermal-hydraulic analysis code ALARM-P1, and by combining newly developed routines for reactor power control system. As ALARM-P1, SHETEMP solves point reactor kinetics equations by the modified Runge-Kutta method and one-dimensional transient heat conduction equations for slab and cylindrical geometries by the Crank-Nicholson methods. The model for reactor power control system takes into account effects of PID regulator and control rod drive mechanism. In order to check errors in programming of the code, calculated results by SHETEMP were compared with analytic solution. Based on the comparisons, the appropriateness of the programming was verified. Also, through a sample calculation for typical modelling, it was concluded that the code could satisfy the fast running capability required for audit and design calculations. This report will be described as a code manual of SHETEMP. It contains descriptions on a sample problem, code structure, input data specifications and usage of the code, in addition to analytical models and results of code verification calculations. (author)

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

    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)

  8. Aircraft Loss-of-Control Accident Analysis

    Belcastro, Christine M.; Foster, John V.

    2010-01-01

    Loss of control remains one of the largest contributors to fatal aircraft accidents worldwide. Aircraft loss-of-control accidents are complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. To gain a better understanding into aircraft loss-of-control events and possible intervention strategies, this paper presents a detailed analysis of loss-of-control accident data (predominantly from Part 121), including worst case combinations of causal and contributing factors and their sequencing. Future potential risks are also considered.

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

    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.)

  10. Safety criteria and guidelines for MSR accident analysis

    Accident analysis for Molten Salt Reactor (MSR) has been investigated at ORNL for MSRE in 1960s. Since then, safety criteria or guidelines have not been defined for MSR accident analysis. Regarding the safety criteria, the authors showed one proposal in this paper. In order to establish guidelines for MSR accident analysis, we have to investigate all possible accidents. In this paper, the authors describe the philosophy for accident analysis, and show 40 possible accidents. They are at first classified as external cause accidents and internal cause accidents. Since the former ones are generic accidents, we investigate only the latter ones, and categorize them to 4 types, such as power excursion accident, flow decrease accident, fuel-salt leak accident, and other accidents mostly specific to MSR. Each accident is described briefly, with some numerical results by the authors. (author)

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

    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.

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

    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

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

    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.

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

    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

  15. Power Excursion Accident Analysis of Research Water Reactor

    A three-dimensional neutronic code POWEX-K has been developed, and it has been coupled with the sub-channel thermal-hydraulic core analysis code SV based on the Single Mass Velocity Model. This forms the integrated neutronic/thermal hydraulics code system POWEX-K/SV for the accident analysis. The Training and Research Reactors at Budapest University of Technology and Economics (BME-Reactor) has been taken as a reference reactor. The cross-section generation procedure based on WIMS. The code uses an implicit difference approach for both the diffusion equations and thermal-hydraulics modules, with reactivity feedback effects due to coolant and fuel temperatures. The code system was applied to analyzing power excursion accidents initiated by ramp reactivity insertion of 1.2 $. The results show that the reactor is inherently safe in case of such accidents i.e. no core melt is expected even if the safety rods do not fall into the core

  16. Reactor accident analysis and evaluation

    Reactor Management Division of Korea Advanced Energy Research Institute has, so far, adopted, modified and developed quite a number of large programs for nuclear core analysis. During the course of this work, it was found necessary to employ some standard subroutines for handling data, input procedures, core memory management and search files. Many programs share lots of common subroutines and/or functions with other programs. Above all, some of them are in lack of transmittal. During the installation of big codes for CYBER computer, it has drawn our keen attention that many elementary subroutines are heavily machine-dependent and that their conversion is extremely difficult. After having collected and modified the subroutines to fit in different codes, it was finally named KINEP (KAERI Improved Nuclear Environmental Package). KINEP has been proved to be convenient even for smaller programs for general purpose. The KINEP includes about one hundred subroutines to facilitate data handling, operator communications, storage allocation, decimal input, file maintence and scratch I/O. (Author)

  17. Interface requirements to couple thermal-hydraulic codes to severe accident codes: ATHLET-CD

    Trambauer, K. [GRS, Garching (Germany)

    1997-07-01

    The system code ATHLET-CD is being developed by GRS in cooperation with IKE and IPSN. Its field of application comprises the whole spectrum of leaks and large breaks, as well as operational and abnormal transients for LWRs and VVERs. At present the analyses cover the in-vessel thermal-hydraulics, the early phases of core degradation, as well as fission products and aerosol release from the core and their transport in the Reactor Coolant System. The aim of the code development is to extend the simulation of core degradation up to failure of the reactor pressure vessel and to cover all physically reasonable accident sequences for western and eastern LWRs including RMBKs. The ATHLET-CD structure is highly modular in order to include a manifold spectrum of models and to offer an optimum basis for further development. The code consists of four general modules to describe the reactor coolant system thermal-hydraulics, the core degradation, the fission product core release, and fission product and aerosol transport. Each general module consists of some basic modules which correspond to the process to be simulated or to its specific purpose. Besides the code structure based on the physical modelling, the code follows four strictly separated steps during the course of a calculation: (1) input of structure, geometrical data, initial and boundary condition, (2) initialization of derived quantities, (3) steady state calculation or input of restart data, and (4) transient calculation. In this paper, the transient solution method is briefly presented and the coupling methods are discussed. Three aspects have to be considered for the coupling of different modules in one code system. First is the conservation of masses and energy in the different subsystems as there are fluid, structures, and fission products and aerosols. Second is the convergence of the numerical solution and stability of the calculation. The third aspect is related to the code performance, and running time.

  18. Development of Evaluation Technology for Hydrogen Combustion in containment and Accident Management Code for CANDU

    Kim, S. B.; Kim, D. H.; Song, Y. M.; and others

    2011-08-15

    For a licensing of nuclear power plant(NPP) construction and operation, the hydrogen combustion and hydrogen mitigation system in the containment is one of the important safety issues. Hydrogen safety and its control for the new NPPs(Shin-Wolsong 1 and 2, Shin-Ulchin 1 and 2) have been evaluated in detail by using the 3-dimensional analysis code GASFLOW. The experimental and computational studies on the hydrogen combustion, and participations of the OEDE/NEA programs such as THAI and ISP-49 secures the resolving capabilities of the hydrogen safety and its control for the domestic nuclear power plants. ISAAC4.0, which has been developed for the assessment of severe accident management at CANDU plants, was already delivered to the regulatory body (KINS) for the assessment of the severe accident management guidelines (SAMG) for Wolsong units 1 to 4, which are scheduled to be submitted to KINS. The models for severe accident management strategy were newly added and the graphic simulator, CAVIAR, was coupled to addition, the ISAAC computer code is anticipated as a platform for the development and maintenance of Wolsong plant risk monitor and Wolsong-specific SAMG.

  19. THYDE-B1/MOD1: a computer code for analysis of small-break loss-of-coolant accident of boiling water reactors

    THYDE-B1/MOD1 is a computer code to analyze thermo-hydraulic transients of the reactor cooling system of a BWR, mainly during a small-break loss-of-coolant accidnet (SB-LOCA) with a special emphasis on the behavior of pressure and mixture level in the pressure vessel. The coolant behavior is simulated with a volume-and-junction method based on assumptions of thermal equilibrium and homogeneous conditions for two-phase flow. A characteristic feature of this code is a three-region representation of the state of the coolant in a control volume, in which three regions, i.e., subcooled liquid, saturated mixture and saturated steam regions are allowed to exist. The regions are separated by moving boundaries, tracked by mass and energy balances for each region. The interior of the pressure vessel is represented by two volumes with three regions: one for inside of the shroud and the other for outside, while other portions of the system are treated with homogeneous model. This method, although it seems to be very simple, has been verified to be adequate for cases of BWR SB-LOCAs in which the hydraulic transient is relatively slow and the cooling of the core strongly depends on the mixture level behavior in the vessel. In order to simulate the system behavior, THYDE-B1 is provided with analytical models for reactor kinetics, heat generation and conduction in fuel rods and structures, heat transfer between coolant and solid surfaces, coolant injection systems, breaks and discharge systems, jet pumps, recirculation pumps, and so on. The verification of the code has been conducted. A good predictability of the code has been indicated through the comparison of calculated results with experimental data provided by ROSA-III small-break tests. This report presents the analytical models, solution method, and input data requirements of the THYDE-B1/MOD1 code. (author)

  20. Modelling of Core Degradation and Progression of Severe Accident by Using MELCOR Code

    After Fukushima Daiichi Nuclear Accident, every single nuclear-field organization in the world focused in the analysis and study of scenarios that leads to core damage and hydrogen releases, in this way the integrated code MELCOR is used by the Mexican Regulatory Body as a tool in the analysis of severe accident progression, core melting and degradation. Scenarios related to core melting could provide information that show important parameters such as: time to reach the core damage, time window for level recovery, etc. This information is useful in the analysis of progression for this kind of events. In this work, Mexican Regulatory Body presents two simulations for different scenarios: a) Station Blackout with no cooling water injection and b) Station Blackout with late cooling water injection. Those two scenarios enclose the response of the fuel under Severe Accident conditions (progression of melting, relocation, temperature profile), plots in this document are qualitative items that allow to analyze the behavior for fuel/core elements. (author)

  1. Accident analysis for the NCSC foil experiment

    An accident analysis has been performed for the nuclear criticality safety class (NCSC) foil experiment. The Los Alamos Critical Experiments Facility (LACEF) performs this experiment regularly during its 2-, 3-, and 5-day nuclear criticality safety classes. This accident analysis is part of an effort to modify the NCSC foil experiment plan so that the experiment may be operated at delayed critical. Currently, the NCSC foil experiment may only be operated up to a neutron multiplication of 100. The purpose of the accident analysis is to ensure that any accidental nuclear excursion does not exceed the boundary of the safety envelope described in the LACEF safety analysis report (SAR). The experiment consists of very thin, highly enriched (93% 235U) uranium metal foils (23 X 23 X 0.008 cm) interleaved between Lucite plates (36 X 36 X 1.27 cm). The fuel foils and Lucite plates are stacked vertically to form a critical assembly. Extra Lucite plates placed at the top and bottom of the assembly act as vertical reflectors. The assembly is operated remotely with the use of a general-purpose vertical-lift platform machine. The accident scenario consists of one additional fuel foil being added to an existing critical or nearly critical stack. The reactivity insertion rate is 0.05 $/s, based on the speed of the vertical-lift platform. It is assumed that none of the safety systems will function properly during the accident and that the operating crew is unable to mitigate the accident

  2. The analysis of thermal-hydraulic models in MELCOR code

    The objective of the present work is to verify the prediction and analysis capability of MELCOR code about the progression of severe accidents in light water reactor and also to evaluate appropriateness of thermal-hydraulic models used in MELCOR code. Comparing the results of experiment and calculation with MELCOR code is carried out to achieve the above objective. Specially, the comparison between the CORA-13 experiment and the MELCOR code calculation was performed

  3. A computer code (WETBERAN) for wet sequence behavior of radioactive nuclides in LWR plant at accident conditions

    The WETBERAN code has been developed to simulate the isotopic- and time-dependent behavior fission products (FP) which leak through the multiple paths of liquid and gas flow within an LWR plant under accident conditions. In this code, emphasis is put on the phenomena pertinent to the presence of water. The TMI, SL-1, and Ginna accidents are analyzed to show the code capability. The TMI 40 day analysis gives detailed informations of FP behavior, both leaking from and remaining in the plant, and proves the effectiveness of the network model for describing the multiple leakage paths. The SL-1 analysis is made to study halogen reduction by water, which cannot be taken into account by CORRAL. The Ginna analysis has been made to check iodine transport by droplets usually generated by primary water flashing at SG tube rupture

  4. Accident analysis in nuclear power plants

    The way the philosophy of Safety in Depth can be verified through the analysis of simulated accidents is shown. This can be achieved by verifying that the integrity of the protection barriers against the release of radioactivity to the environment is preserved even during accident conditions. The simulation of LOCA is focalized as an example, including a study about the associated environmental radiological consequences. (Author)

  5. A thermo mechanical benchmark calculation of a hexagonal can in the BTI accident with INCA code

    The thermomechanical behaviour of an hexagonal can in a benchmark problem (simulating the conditions of a BTI accident in a fuel assembly) is examined by means of the INCA code and the results systematically compared with those of ADINA

  6. Long term cooling analysis after Fukushima Daiichi accident

    The objective of this study is to analyze of the long term cooling after Fukushima Daiichi accident by RELAP5mode3.3 code and to check the validity of the cooling method. In order to simulate the cooling conditions in Fukushima plants after accident, the model is nodalized on the assumption of the existence of steam/liquid leak position from RPV/PCV and the variety of debris distribution in RPV/PCV. As a result, we estimated the debris distribution in RPV by referring plant parameter such as reactor pressure and temperature. In addition, we performed the analysis of the loss of injection water accident for the current cooling system installed in Fukushima Daiichi cite after the earthquake. In this case, we develop simplified nodalization of RPV to analyze temperature behavior of reactor structural materials by using the radiation heat transfer model. (author)

  7. Analysis on the severe accidents in KSTAR tokamak

    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.

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

    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. A study on the core analysis methodology for SMART CEA ejection accident-I

    A methodology to analyze the fuel enthalpy is developed based on MASTER that is a time dependent 3 dimensional core analysis code. Using the proposed methodology, SMART CEA ejection accident is analyzed. Moreover, radiation doses are estimated at the exclusion area boundary and low population zone to confirm the criteria for the accident. (Author). 31 refs., 13 tabs., 18 figs

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

    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

  11. Safety analysis of surface haulage accidents

    Randolph, R.F.; Boldt, C.M.K.

    1996-12-31

    Research on improving haulage truck safety, started by the U.S. Bureau of Mines, is being continued by its successors. This paper reports the orientation of the renewed research efforts, beginning with an update on accident data analysis, the role of multiple causes in these accidents, and the search for practical methods for addressing the most important causes. Fatal haulage accidents most often involve loss of control or collisions caused by a variety of factors. Lost-time injuries most often involve sprains or strains to the back or multiple body areas, which can often be attributed to rough roads and the shocks of loading and unloading. Research to reduce these accidents includes improved warning systems, shock isolation for drivers, encouraging seatbelt usage, and general improvements to system and task design.

  12. Strategy for the Development of Severe Accident Analysis Technology

    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

  13. International activities for the analysis of the TMI-2 accident with special consideration of ATHLET calculations

    Several OECD countries still have great interest to analyze the TMI-2 accident. Thermal hydraulic best estimate codes and severe accident codes are used to calculate the TMI-2 analysis exercise defined by a CSNI task group. Fourteen organizations in nine OECD countries are participating in the exercise. Four thermal hydraulic best estimate codes and six severe accident codes are used. The Federal Republic of Germany (FRG) is using the thermal hydraulic code ATHLET developed in the GRS to calculate the TMI-2 analysis exercise. Lessons learned are concentrated on the assessment of ATHLET, show advantages of the two phase thermal hydraulic model used, and identify areas for further development. Results from ATHLET calculations are compared with results from other OECD-codes. (orig.)

  14. Development of methods for the analysis of accident scenarios with steam line breaks and boron dilution by the help of the code system ATHLET-DYN3D. Final report. Pt. 1

    Libraries of two-group neutron-diffusion parameters for a Siemens-KWU-Konvoi Pressurized Water Reactor have been generated at Forschungszentrum Rossendorf and TUeV Bau und Betrieb GmbH by using the codes HELIOS and CASMO, respectively. The libraries have been coupled to the reactor-dynamics code DYN3D. For a generic PWR core containing MOX fuel elements, DYN3D macro-burnup calculations and the calculation of different operation states have been carried out. The results will be used for the investigation of possible accident scenarios. Reactivity coefficients calculated by DYN3D are needed for accident analyses by the 1-D thermal-hydraulic code ATHLET. Using the cross section data, more detailed analyses can be carried out by applying the coupled-code system DYN3D-ATHLET, considering 3D neutron kinetics. The comparison of the results calculated by DYN3D with two different diffusion-parameter libraries can give an idea of how uncertainties in diffusion data influence the accuracy of reactor simulation. (orig.)

  15. Analytical validation of the CACECO containment analysis code

    The CACECO containment analysis code was developed to predict the thermodynamic responses of LMFBR containment facilities to a variety of accidents. This report covers the verification of the CACECO code by problems that can be solved by hand calculations or by reference to textbook and literature examples. The verification concentrates on the accuracy of the material and energy balances maintained by the code and on the independence of the four cells analyzed by the code so that the user can be assured that the code analyses are numerically correct and independent of the organization of the input data submitted to the code

  16. Investigation of NPP behavior in case of loss of coolant accident based on comparison of different ASTEC computer code versions

    The paper presents the work performed at the Institute for Nuclear Research and Nuclear Energy (INRNE) and Bhabha Atomic Research Centre (BARC), India in the frame of SARNET2 project. The performed work continues the effort in the field of nuclear safety and cooperation between INRNE-BAS and BARC. The main target is development and validation of ASTEC (Accident Source Term Evaluation Code) at the further, a tool for level-2 PSA analysis for better understanding of accident progression during in-vessel phase until reactor vessel failure. (authors)

  17. Code Coupling for Multi-Dimensional Core Transient Analysis

    Park, Jin-Woo; Park, Guen-Tae; Park, Min-Ho; Ryu, Seok-Hee; Um, Kil-Sup; Lee Jae-Il [KEPCO NF, Daejeon (Korea, Republic of)

    2015-05-15

    After the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the Doppler effect becomes important and turns the reactivity balance and power down to lower levels. Although this happens in a very short period of time, only few seconds, the energy generated can be very significant and cause fuel failures. The current safety analysis methodology which is based on overly conservative assumptions with the point kinetics model results in quite adverse consequences. Thus, KEPCO Nuclear Fuel(KNF) is developing the multi-dimensional safety analysis methodology to mitigate the consequences of the single CEA ejection accident. For this purpose, three-dimensional core neutron kinetics code ASTRA, sub-channel analysis code THALES, and fuel performance analysis code FROST, which have transient calculation performance, were coupled using message passing interface (MPI). This paper presents the methodology used for code coupling and the preliminary simulation results with the coupled code system (CHASER). Multi-dimensional core transient analysis code system, CHASER, has been developed and it was applied to simulate a single CEA ejection accident. CHASER gave a good prediction of multi-dimensional core transient behaviors during transient. In the near future, the multi-dimension CEA ejection analysis methodology using CHASER is planning to be developed. CHASER is expected to be a useful tool to gain safety margin for reactivity initiated accidents (RIAs), such as a single CEA ejection accident.

  18. Comparison of Severe Accident Results Among SCDAP/RELAP5, MAAP, and MELCOR Codes

    This paper demonstrates a large-break loss-of-coolant accident (LOCA) sequence of the Kuosheng nuclear power plant (NPP) and station blackout sequence of the Maanshan NPP with the SCDAP/RELAP5 (SR5), Modular Accident Analysis Program (MAAP), and MELCOR codes. The large-break sequence initiated with double-ended rupture of a recirculation loop. The main steam isolation valves (MSIVs) closed, the feedwater pump tripped, the reactor scrammed, and the assumed high-pressure and low-pressure spray systems of the emergency core cooling system (ECCS) were not functional. Therefore, all coolant systems to quench the core were lost. MAAP predicts a longer vessel failure time, and MELCOR predicts a shorter vessel failure time for the large-break LOCA sequence. The station blackout sequence initiated with a loss of all alternating-current (ac) power. The MSIVs closed, the feedwater pump tripped, and the reactor scrammed. The motor-driven auxiliary feedwater system and the high-pressure and low-pressure injection systems of the ECCS were lost because of the loss of all ac power. It was also assumed that the turbine-driven auxiliary feedwater pump was not functional. Therefore, the coolant system to quench the core was also lost. MAAP predicts a longer time of steam generator dryout, time interval between top of active fuel and bottom of active fuel, and vessel failure time than those of the SR5 and MELCOR predictions for the station blackout sequence. The three codes give similar results for important phenomena during the accidents, including SG dryout, core uncovery, cladding oxidation, cladding failure, molten pool formulation, debris relocation to the lower plenum, and vessel head failure. This paper successfully demonstrates the large-break LOCA sequence of the Kuosheng NPP and the station blackout sequence of the Maanshan NPP

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

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

  20. Modeling of a confinement bypass accident with CONSEN, a fast-running code for safety analyses in fusion reactors

    Highlights: • The CONSEN code for thermal-hydraulic transients in fusion plants is introduced. • A magnet induced confinement bypass accident in ITER has been simulated. • A comparison with previous MELCOR results for the accident is presented. -- Abstract: The CONSEN (CONServation of ENergy) code is a fast running code to simulate thermal-hydraulic transients, specifically developed for fusion reactors. In order to demonstrate CONSEN capabilities, the paper deals with the accident analysis of the magnet induced confinement bypass for ITER design 1996. During a plasma pulse, a poloidal field magnet experiences an over-voltage condition or an electrical insulation fault that results in two intense electrical arcs. It is assumed that this event produces two one square meters ruptures, resulting in a pathway that connects the interior of the vacuum vessel to the cryostat air space room. The rupture results also in a break of a single cooling channel within the wall of the vacuum vessel and a breach of the magnet cooling line, causing the blow down of a steam/water mixture in the vacuum vessel and in the cryostat and the release of 4 K helium into the cryostat. In the meantime, all the magnet coils are discharged through the magnet protection system actuation. This postulated event creates the simultaneous failure of two radioactive confinement barrier and it envelopes all type of smaller LOCAs into the cryostat. Ice formation on the cryogenic walls is also involved. The accident has been simulated with the CONSEN code up to 32 h. The accident evolution and the phenomena involved are discussed in the paper and the results are compared with available results obtained using the MELCOR code

  1. Modelling of severe accident behaviour using the code ATHLET-CD

    Thermal-hydraulic and core degradation phenomena play a decisive role for the course of severe accidents in light water reactors. Therefore, the simulation of such accidents with computer codes requires comprehensive and detailed modelling of these processes. The code ATHLET-CD is being developed for realistic simulation of accidents with core degradation and for evaluation of accident management measures. It makes use of the detailed and validated models of the thermal-hydraulic code ATHLET in an efficient coupling with models for core degradation and fission product behaviour. The capabilities of the coupled code are demonstrated by means of the calculation of the TMI-2 accident. The first three phases of the accident were successfully simulated in a reasonable computing time. The calculated system pressure and pressurizer level after pump trip, during the pump restart, and until core slump are in acceptable agreement with the measured data. The calculated hydrogen generation before the pump restart is in accordance with the deduced value. Contrary to estimates based on the system behaviour, no significant hydrogen generation was calculated during the quench phase. Further model improvements regarding the quenching of degraded core material, fracture and relocation of solid fuel rods, as well as the simulation of debris bed behaviour are necessary for better simulation. (authors)

  2. Lessons learnt from the EC/USNRC expert judgement study on probabilistic accident consequence codes applied in the COSYMA uncertainty analyses

    Two probabilistic accident consequence codes, COSYMA and MACCS respectively, estimate the risks and other endpoints associated with hypothetical accidents from nuclear installations. A joint EC/USNRC project for an uncertainty analysis of these two codes was initiated to systematically derive credible and traceable probability distributions for the respective code input variables. A formal expert judgement elicitation and evaluation process was used as the best available technique to accomplish that objective. These input distributions were used in an uncertainty analysis of the COSYMA package. This paper will show the overall process and highlights the lessons learnt from the projects. (author)

  3. Fuel safety analysis following feeder break accident for refurbished Wolsong 1

    The objective of the fuel analysis for the postulated accident was to estimate the quantity and timing of a fission product release from fuels when a postulated single channel accident occurs in CANDU 6 reactors. In this study, a fuel safety analysis for the refurbished Wolsong 1 was carried out by using the latest IST (Industrial Standard Toolset) fuel code. The relevant accident scenario focused in this study was a feeder stagnation break accident. The amount of fission product inventory and its distribution during the normal operating conditions were calculated by using the latest ELESTRES-IST code. For a calculation of transient fission product release following the feeder stagnation break, it was assumed that all fuel sheaths in the channel were failed and the entire gap inventory was released instantaneously at the beginning of the accident. The additional releases from the grain boundary and in-grain bound inventories were estimated by applying the Gehl's release model. (author)

  4. User's manual of ART code for analyzing fission product transport behavior during core meltdown accident

    In a probabilistic risk assessment (PRA) it has been recognized that a core meltdown accident with a large amount of fission products released to the environment is a dominant contributor to public risk. For the evaluation of the risk, information about source terms are inevitable. In order to analyze fission product transport behavior and to evaluate source terms during a core meltdown accident, the ART code has been developed. The ART code has the following features: (1) It can treat fission product transport behavior both in a primary system and a containment system, (2) It models fission product transport caused by both gas flow and liquid flow, and (3) It includes a detailed model about transport behavior of aerosols which are released in quantity during a core meltdown accident. This report is a user's manual for the ART code and includes description of modeling, input/output data and a sample run. (author)

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

    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

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

    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.

  7. Suppression pool swell analysis using CFD code

    A two-dimensional axi-symmetric model of suppression pool of Containment Studies Facility (CSF) along with single vent pipe was modeled to estimate the jet and hydrodynamic loads due to flow of steam air mixture during simulated loss of coolant accident (LOCA). The analysis was carried out using CFD ACE+ software with Volume of Fluid (VOF) approach. The flow velocity variation through vent pipe was estimated using in-house containment thermal hydraulic code CONTRAN, was given as input at inlet boundary condition. The transient calculations were performed for 20 seconds and suppression pool level variation, pressure loads over the floor, walls and vent pipes etc were evaluated. (author)

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

    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

  9. Analysis of reactivity induced accidents at Pakistan Research Reactor-1

    Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m3/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure

  10. Analysis of reactivity induced accidents at Pakistan Research Reactor-1

    Bokhari, I.H. E-mail: ishtiaq@pinstech.org.pk; Israr, M.; Pervez, S

    2002-12-01

    Analysis of reactivity induced accidents in Pakistan Research Reactor-1 (PARR-1) utilizing low enriched uranium (LEU) fuel, has been carried out using standard computer code PARET. The present core comprises of 29 standard and five control fuel elements. Various modes of reactivity insertions have been considered. The events studied include: start-up accident; accidental drop of a fuel element on the core; flooding of a beam tube with water; removal of an in-pile experiment during reactor operation etc. For each of these transients, time histories of reactor power, energy released and clad surface temperature etc. were calculated. The results reveal that the peak clad temperatures remain well below the clad melting temperature during these accidents. It is concluded that the reactor, which is operated safely at a steady-state power level of 10 MW, with coolant flow rate of 950 m{sup 3}/h, will also be safe against any possible reactivity induced accident and will not result in a fuel failure.

  11. Analysis of reactivity insertion accidents in PWR reactors

    A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author)

  12. On the application of near accident data to risk analysis of major accidents

    Major accidents are low frequency high consequence events which are not well supported by conventional statistical methods due to data scarcity. In the absence or shortage of major accident direct data, the use of partially related data of near accidentsaccident precursor data – has drawn much attention. In the present work, a methodology has been proposed based on hierarchical Bayesian analysis and accident precursor data to risk analysis of major accidents. While hierarchical Bayesian analysis facilitates incorporation of generic data into the analysis, the dependency and interaction between accident and near accident data can be encoded via a multinomial likelihood function. We applied the proposed methodology to risk analysis of offshore blowouts and demonstrated its outperformance compared to conventional approaches. - Highlights: • Probabilistic risk analysis is applied to model major accidents. • Two-stage Bayesian updating is used to generate informative distributions. • Accident precursor data are used to develop likelihood function. • A multinomial likelihood function is introduced to model dependencies among data

  13. Thermohydraulic and Safety Analysis for CARR Under Station Blackout Accident

    A thermohydraulic and safety analysis code (TSACC) has been developed using Fortran 90 language to evaluate the transient thermohydraulic behaviors and safety characteristics of the China Advanced Research Reactor(CARR) under Station Blackout Accident(SBA). For the development of TSACC, a series of corresponding mathematical and physical models were considered. Point reactor neutron kinetics model was adopted for solving reactor power. All possible flow and heat transfer conditions under station blackout accident were considered and the optional models were supplied. The usual Finite Difference Method (FDM) was abandoned and a new model was adopted to evaluate the temperature field of core plate type fuel element. A new simple and convenient equation was proposed for the resolution of the transient behaviors of the main pump instead of the complicated four-quadrant model. Gear method and Adams method were adopted alternately for a better solution to the stiff differential equations describing the dynamic behaviors of the CARR. The computational result of TSACC showed the enough safety margin of CARR under SBA. For the purpose of Verification and Validation (V and V), the simulated results of TSACC were compared with those of Relap5/Mdo3. The V and V result indicated a good agreement between the results by the two codes. Because of the adoption of modular programming techniques, this analysis code is expected to be applied to other reactors by easily modifying the corresponding function modules. (authors)

  14. Calculation and Analysis of Neutron Time-spatial Kinetics in uncontrolled withdrawal accident of regulating rod in CEFR

    In the accident analysis of China Experimental Fast Reactor (CEFR), uncontrolled withdrawal of regulating rod without scram is a severe accident. When it happens, a large positive reactivity will be introduced and the relative space distribution of neutron flux will change significantly. Thus the analytical result of the accident by point kinetics in the safety analysis of CEFR is inaccurate. In this paper, focused on neutronics behavior, the accident is re-evaluated with NAS-K code. The NAS-K code is a CEFR-self-developed three-dimensional space-time-dependent neutron kinetics code for sodium cooled fast reactors, including thermal feedback and various kinds of reactivity feedback effects. The calculation results indicate that the maximum temperatures of fuel and cladding do not exceed the limits specified by acceptance criteria corresponding to design basis accident for CEFR, which means the accident will not cause damage to CEFR core. (author)

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

    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

  16. Simulation of rod ejection accident byPARCS code

    Matějková, J.

    2015-01-01

    This paper describes reactor core model used for simulating REA. The model was designed in PARCS utilizing graphical interface SNAP. The data for model were given from benchmark NEACPR L-335. The PARCS model used integrated thermal hydraulic block for calculation. The results and solution is shown in the paper. Thermal hydraulic calculation can also be provided by external system code TRACE. The PARCS model is prepared to couple with TRACE model for giving more accurate calculation.

  17. Adaptation of the severe accident codes to VVER-440/V213 (V230) reactor unit, their comparison and utilisation of the results

    This paper presents an application and comparison of the computer codes, devoted for severe accident analysis of PWR up to source term evaluation, to a VVER-440/V213 and V230 NPP. The basic results of selected sequences are described and some physical parameters predicted by different codes are compared. The comparison is deliberated mainly on the timing of main primary circuit events and fission products behaviour up to source term evaluation. Utilisation of the results of the severe accident analysis for development of the emergency procedures for rapid assessment of barriers status and source term category is shortly described, too. (author)

  18. ENSDF ANALYSIS AND UTILITY CODES.

    BURROWS, T.

    2005-04-04

    The ENSDF analysis and checking codes are briefly described, along with their uses with various types of ENSDF datasets. For more information on the programs see ''Read Me'' entries and other documentation associated with each code.

  19. PWR Core 2 Project accident analysis

    The various operations required for receipt, handling, defueling and storage of spent Shippingport PWR Core 2 fuel assemblies have been evaluated to determine the potential accidents and their consequences. These operations will introduce approximately 16,500 kilograms of depleted natural uranium (as UO2), 139 kilograms of plutonium, 2.8 megacuries of mixed fission products, and 14 kilograms of Zircaloy-4 (cladding and hardware) into the 221-T Canyon Building. Event sequences for potential accidents that were considered included (1) leaking fuel assemblies, (2) fire and explosion, (3) loss of coolant or cooling capability, (4) dropped and/or damaged fuel assemblies, and extrinsic occurrences such as loss of services, missile impact, and natural occurrences (e.g., earthquake, tornado). Accident frequencies were determined by formal analysis to be very low. Accident consequences are greatly mitigated by the safety and containment features designed into the fuel modules and shipping cask, the long cooling time since reactor discharge, and the redundant safety features designed into the facilities, equipment, and operating procedures for the PWR Core 2 Project. Possible hazards associated with the handling of these fuels have been considered and adequate safeguards and storage constraints identified. The operations of M-160 cask unloading and module storage will not involve identifiable risks as great or significantly greater than those for comparable licensed nuclear facilities, nor will hazards or risks be significantly different from comparable past 221-T Plant programs. Therefore, it is concluded that the operations required for receipt, handling, and defueling of the M-160 cask and for the storage and surveillance of the PWR Core 2 fuel assemblies at the 221-T Canyon Building can be performed without undue risk to the safety of the involved personnel, the public, the environment or the facility

  20. Quality assurance and verification of the MACCS [MELCOR Accident Consequence Code System] code, Version 1.5

    An independent quality assurance (QA) and verification of Version 1.5 of the MELCOR Accident Consequence Code System (MACCS) was performed. The QA and verification involved examination of the code and associated documentation for consistent and correct implementation of the models in an error-free FORTRAN computer code. The QA and verification was not intended to determine either the adequacy or appropriateness of the models that are used MACCS 1.5. The reviews uncovered errors which were fixed by the SNL MACCS code development staff prior to the release of MACCS 1.5. Some difficulties related to documentation improvement and code restructuring are also presented. The QA and verification process concluded that Version 1.5 of the MACCS code, within the scope and limitations process concluded that Version 1.5 of the MACCS code, within the scope and limitations of the models implemented in the code is essentially error free and ready for widespread use. 15 refs., 11 tabs

  1. Calculations of reactor-accident consequences, Version 2. CRAC2: computer code user's guide

    The CRAC2 computer code is a revision of the Calculation of Reactor Accident Consequences computer code, CRAC, developed for the Reactor Safety Study. The CRAC2 computer code incorporates significant modeling improvements in the areas of weather sequence sampling and emergency response, and refinements to the plume rise, atmospheric dispersion, and wet deposition models. New output capabilities have also been added. This guide is to facilitate the informed and intelligent use of CRAC2. It includes descriptions of the input data, the output results, the file structures, control information, and five sample problems

  2. Initial event analysis of the Fukushima Daiichi accident

    The objective of this study is to investigate the initial event of the Fukushima Daiichi accident and to check the validity of the counter measures against the accident. We analyzed the initial event of the Fukushima Daiichi accident for Unit-1, 2 and 3 plants by RELAP5 code and compared with the actual plant data. The parametric study about the operation of the isolation condenser (IC), the reactor core isolation cooling system (RCIC) and the high pressure core injection system (HPCI) was also done to understand the accident progression. (author)

  3. Analysis of progression of severe accident in Indian PHWRs

    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

  4. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    This document provides the detailed accident analysis to support ''HNF-3553, Spent Nuclear Fuel Project Final Safety, Analysis Report, Annex A,'' ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  5. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    CROWE, R.D.; PIEPHO, M.G.

    2000-03-23

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  6. Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

    This document provided the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report''. All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  7. Canister storage building design basis accident analysis documentation

    KOPELIC, S.D.

    1999-02-25

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report.

  8. Canister storage building design basis accident analysis documentation

    This document provides the detailed accident analysis to support HNF-3553, Spent Nuclear Fuel Project Final Safety Analysis Report, Annex A, ''Canister Storage Building Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the Canister Storage Building Final Safety Analysis Report

  9. Core disruptive accident analysis using ASTERIA-FBR

    Japan Nuclear Energy Safety Organization (JNES) is developing a core disruptive accident analysis code, ASTERIA-FBR, which tightly couples the thermal-hydraulics and the neutronics to simulate the core behavior during core disruptive accidents (CDA) of fast breeder reactors (FBRs). ASTERIA-FBR consists of the three-dimensional thermal-hydraulics calculation module: CONCORD, the fuel pin behavior calculation module: FEMAXI-FBR, and the space-time neutronics module: Dynamic-GMVP or PARTISN/RKIN. This paper describes a comparison between characteristics of GMVP and PARTISN and summarizes the challenging issues on applying Dynamic-GMVP to the calculation against unprotected loss-of-flow (ULOF) event which is a typical initiator of core disruptive accident of FBR. It was found that Dynamic-GMVP is confirmed to be basically applicable to the CDA phenomena. It was found that, however, applying GMVP to the CDA calculation is less reasonable than PARTISN since the calculation load of GMVP is too large to meet the required calculation accuracy, although the Monte-Carlo method is based on the actual neutron behavior without any discretization of space and energy. The statistical error included in the calculation results may affect the super-prompt criticality during ULOF event and thus the amount of released energy

  10. RSM modelling of an ATWS accident simulated by the ALMOD code: methodological and practical achievement

    A simulation study of a PWR station black-out ATWS has been performed by applying Response Surface Methodology (RSM) on the data obtained by inspecting the ALMOD code. The case under study has shown that the a priori information which alone could be inadequate, is optimally utilized if coupled with a preliminary sensitivity analysis through RSM techniques. In particular the engineering selection of the model variables and the rank order of the remaining ones had to be modified after an RSM preliminary sensitivity analysis. An other qualifying feature of the exercise is the use of randomization of the variables not included in the model in order to coherently exploit the methodology in its full efficiency. This procedure is able to give a figure of merit of the global importance of the neglected variables through the analysis of residuals. Results show that the proposed technique is an effective tool for selecting the most important accident variables and that the body of information gained is significant with respect to the number of observations performed

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

    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. Fuel performance analysis code 'FAIR'

    For modelling nuclear reactor fuel rod behaviour of water cooled reactors under severe power maneuvering and high burnups, a mechanistic fuel performance analysis code FAIR has been developed. The code incorporates finite element based thermomechanical module, physically based fission gas release module and relevant models for modelling fuel related phenomena, such as, pellet cracking, densification and swelling, radial flux redistribution across the pellet due to the build up of plutonium near the pellet surface, pellet clad mechanical interaction/stress corrosion cracking (PCMI/SSC) failure of sheath etc. The code follows the established principles of fuel rod analysis programmes, such as coupling of thermal and mechanical solutions along with the fission gas release calculations, analysing different axial segments of fuel rod simultaneously, providing means for performing local analysis such as clad ridging analysis etc. The modular nature of the code offers flexibility in affecting modifications easily to the code for modelling MOX fuels and thorium based fuels. For performing analysis of fuel rods subjected to very long power histories within a reasonable amount of time, the code has been parallelised and is commissioned on the ANUPAM parallel processing system developed at Bhabha Atomic Research Centre (BARC). (author). 37 refs

  13. Hindsight Bias in Cause Analysis of Accident

    Atsuo Murata; Yasunari Matsushita

    2014-01-01

    It is suggested that hindsight becomes an obstacle to the objective investigation of an accident, and that the proper countermeasures for the prevention of such an accident is impossible if we view the accident with hindsight. Therefore, it is important for organizational managers to prevent hindsight from occurring so that hindsight does not hinder objective and proper measures to be taken and this does not lead to a serious accident. In this study, a basic phenomenon potentially related to accidents, that is, hindsight was taken up, and an attempt was made to explore the phenomenon in order to get basically insights into the prevention of accidents caused by such a cognitive bias.

  14. Severe accident containment-response and source term analyses by AZORES code for a typical FBR plant

    Japan Nuclear Energy Safety organization (JNES) is developing severe accident analysis codes in order to apply to the probabilistic safety assessment (PSA) for a typical fast breeder reactor (FBR). The AZORES code analyzes the severe accident phenomena in the reactor containment that reactor coolant (sodium) and molten core debris are released from the primary cooling system boundary and the release fraction to the environment of fission products (FP). This report summarized results analyzed using the AZORES code for a PLOHS (loss of decay heat removal function) accident sequence with the actual plant system about the containment bypass (CVBP) scenario, and the containment failure scenario due to hydrogen deflagration or detonation. The results showed that the coolant temperature of the primary system and the secondary system in the PLOHS sequence increased at the almost same temperature, and the creep damage to the reactor coolant boundary became significant when coolant temperature exceeded about 1,100 K. The release fractions of FP in the CVBP case were estimated to be 0.99 for Xe, 0.14 for iodine, 0.44 for Cs and 0.01 for non-volatile tetravalent Ce. The release fractions of FP in the containment vessel failure case due to hydrogen burning were estimated to be 0.82 for Xe, 0.06 for iodine, 0.06 for Cs and 0.003 for non-volatile tetravalent Ce. In the present study, release fractions of FPs to the environment were obtained for the CVBP and the containment failure cases of the PLOHS accident sequence for the typical FBR plant. (author)

  15. Parameterization of the driving time in the evacuation or fast relocation model of an accident consequence code

    The model of protective measures in the accident consequence code system UFOMOD of the German Risk Study, Phase B, requires the driving times of the population to be evacuated for the evaluation of the dose received during the evacuation. The parameter values are derived from evacuation simulations carried out with the code EVAS for 36 sectors from various sites. The simulations indicated that the driving time strongly depends on the population density, whereas other influences are less important. It was decided to use different driving times in the consequence code for each of four population density classes as well as for each of three or four fractions of the population in a sector. The variability between sectors of a class was estimated from the 36 sectors, in order to derive subjective probability distributions that are to model the uncertainty in the parameter value to be used for any of the fractions in a particular sector for which an EVAS simulation has not yet been performed. To this end also the impact of the uncertainties in the parameters and modelling assumptions of EVAS on the simulated times was quantified using expert judgement. The distributions permit the derivation of a set of driving times to be used as so-called ''best estimate'' or reference values in the accident consequence code. Additionally they are directly applicable in an uncertainty and sensitivity analysis

  16. Analysis on the Direct Vessel Injection Line Break Accident at APR+ Standard Design

    Lee, Youngho; Yang, Huichang [TUEV Rheinland Korea Ltd., Seoul (Korea, Republic of); Kim, Kap [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2013-10-15

    APR+ (Advanced Power Reactor +) is the newest design variation of APR1400. The main characteristics of APR+, compared with APR1400, are passive safety systems and dedicated systems for severe accident mitigation. APR+ is under review for standard design certification. In this study, thermal hydraulic analysis on the Direct Vessel Injection (DVI) line break accident postulated in APR+ design was performed. Comparisons of the major parameters which can represent the overall accident behavior during DVI line break accident, several discrepancies between this study and reference data were found and such discrepancies include actuation timing of SIPs and SITs, and also include parameter behaviors of break flow rate and PCT at the accident initiation. These differences were mainly from the different thermal hydraulic models in simulation codes. The behavioral differences for break flow as well as peak cladding temperatures will be examined further as a next step for this study.

  17. Cold Vacuum Drying (CVD) Facility Design Basis Accident Analysis Documentation

    PIEPHO, M.G.

    1999-10-20

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR.

  18. Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

    This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report, ''Cold Vacuum Drying Facility Final Safety Analysis Report (FSAR).'' All assumptions, parameters and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR

  19. Methods and codes for assessing the off-site Consequences of nuclear accidents. Volume 2

    The Commission of the European Communities, within the framework of its 1980-84 radiation protection research programme, initiated a two-year project in 1983 entitled methods for assessing the radiological impact of accidents (Maria). This project was continued in a substantially enlarged form within the 1985-89 research programme. The main objectives of the project were, firstly, to develop a new probabilistic accident consequence code that was modular, incorporated the best features of those codes already in use, could be readily modified to take account of new data and model developments and would be broadly applicable within the EC; secondly, to acquire a better understanding of the limitations of current models and to develop more rigorous approaches where necessary; and, thirdly, to quantify the uncertainties associated with the model predictions. This research led to the development of the accident consequence code Cosyma (COde System from MAria), which will be made generally available later in 1990. The numerous and diverse studies that have been undertaken in support of this development are summarized in this paper, together with indications of where further effort might be most profitably directed. Consideration is also given to related research directed towards the development of real-time decision support systems for use in off-site emergency management

  20. Loss of Off Site Power Accident Analysis Probabilistic and Deterministic Approach

    The accident analysis of a reactor should consider analysis of Design Base Accidents DBA . One of these accident is the Loss of Off Site Power LOSP. For Egypt, the LOSP frequency is abnormally high; 10 times/year, this gives a good cause for initiating the present work. The present work adopted both the probabilistic and the deterministic methods to determine the likelihood of the accident and to establish and assertion of the variation of the reactor safety related parameters e.g. power, flow and temperature over the prescribed accident evolution time. The accident scenario covers the failure sequences of the Reactor Safety systems. Four generated LOSP accident scenarios are analyzed. Two codes are used, the first is the IAEA Probabilistic Safety Assessment Package PSAPACK, the second TR22M21 has been developed by the author to simulate the expected behavior of the reactor thermal-hydraulic parameters. It has been found that the two codes have successfully identify a severe scenario with annual occurrences frequency of 3.8 E-5 which could significantly contribute to the risk of the core damage

  1. User's handbook for the iodine severe accident behavior code IMPAIR 2.2

    This publication describes the second version of the Iodine Severe Accident Code (IMPAIR 2.2). This code aims to model postulated conditions of iodine chemistry present in a containment (sump, deposition and atmosphere) during a postulated severe accident in a LWR by using 29 differential equations and 69 rate constants. These equations model the behavior of various iodine species in the sump and the gas phase. Apart from purely chemical equilibria, the mass transport of aerosols, elemental iodine, organoiodine species and droplet carry-over during pressure release while venting are also described. Various improvements and extensions have been made since the original publication. Meanwhile the multi-compartment version, IMPAIR 2/M has become available. It will be updated with the revised models described here as well as other changes. The updated multi-compartment code will be designated IMPAIR 3 and will replace the single-compartment code IMPAIR 2.2. It should be available in 1992. The revised code IMPAIR 2.2, although it has not reached the maturity of an 'assessed code', is significantly improved in its function to model the iodine chemistry of severe accident scenarios in a containment using a mainly phenomenological approach. This improvement was achieved largely through data available from ACE/RTF tests. Two sets of test data with drastic pH difference were used to validate the updated code. The calculated results show a very good correlation for all iodine species in the gas and water phases and in deposition to within an order of magnitude. These validation results were presented at the Third CSNI Workshop on Iodine Chemistry, Tokai-Mura, Japan. (author) 15 tabs., 19 refs

  2. Dispersion of radioactive materials from JRTR following a postulated accident using HOTSPOT code

    Jordan Research and Training Reactor (JRTR) is the first nuclear facility in Jordan. The JRTR is 5 MW, light water moderated and open type pool reactor. In case of an accident, the radioactive materials will be released to the surrounding environment and endanger the people living in the vicinity of the reactor. However, up to now, no study has been published about the dispersion of radioactive materials from JRTR in case of an accident. As preliminary stage for the construction of the JRTR, the dispersion of the radioactive materials from JRTR in case of an accident was studied using HOTSOT code. The result of the report indicates that for ground level release with an average speed of 3.6 m/s of hourly averaged meteorological data for one year with a dominant direction from the west a person located at distance .062 km from the reactor site will receive .25 Sv

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

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

    2010-07-01

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

  4. NSR-77: a computer code for transient analysis of a light water reactor fuel rod

    This report describes computer code NSR-77 written in FORTRAN IV for FACOM-M 200 computer in detail. It has been developed for transient response analysis of a light water reactor fuel rod during an accident such as a reactivityy initiated accident, a loss-of-coolant accident or a power-cooling-mismatch accident. The code consists of subcodes which calculate heat conduction in a fuel rod, gas gap conductance between fuel and cladding, heat transfer from cladding to coolant, fluid hydrodynamics, elastic-plastic fuel and cladding deformation, and material properties, and so on. (author)

  5. Neutronic static analysis of Chernobyl accident

    In the present analysis, estimates were made of the positive reactivity introduced through the growth of the coolant void fraction in a Graphite-water steam-generating reactor both at the average value of burnup given by the Soviets and at the maximum value. Using Monte Carlo models, various possible axial distribution of burnup, displacer models, conditions in the control channels and positions of the control rods were considered in calculating the insertion of positive reactivity with the fall of the manual and emergency control rods; that is the positive scram. The possibility of positive reactivity insertion due to the creation of a mixture of fuel, water and cladding in a number of central fuel channels has been examined. This situation corresponds to the explosion of these channels, and is considered in the present work as the cause of the second reactivity peak. At the level of the data presented in this study, vaporization of cooling water in the fuel channels can be considered as the cause of the Chernobyl accident. The accident began in the region of the channels close to the axis of the reactor and spread to its periphery. The positive reactivity due to insertion of the manual and emergency control rods - positive scram -played a marginal role in the development of the accident. Fracture of the fuel followed by bursting of the channels around the axis of the reactor, due to contact between the hot UO2 particles and the cooling water at th end of the first peak, could have started a mechanism capable of producing a second peak in reactivity, in the case of fuel damage extended to a sufficiently large portion of the core

  6. An analysis of the Three Mile Island accident

    Starting with a systematic analysis of the chain of events that took place during the Three Mile Island accident, the authors assess the significance of the four distinct phases of the accident. Inferences that can be drawn with respect to the safety of CANDU reactors are discussed. A rational reaction to the accident is suggested, and several factors are shown not to have played an important part, contrary to public impressions. The authors point out that over-reaction to the accident could detract from public safety. The Canadian response to the accident is discussed. (auth)

  7. Analysis of the TMI-2 accident using ATHLET-CD

    On March 28, 1979, the only loss-of-coolant accident so far in a light water reactor exceeding the design basis occurred in unit 2 of the U.S. PWR of Three Mile Island (TMI-2) near Harrisburg. The accident entailed massive core degradation accompanied by the formation of a bed of debris and a molten pool. Approx. 30 t of the core inventory were moved to the bottom of the reactor pressure vessel; the vessel sustained this thermal load. Also because of the worldwide use of light water reactors, this accident constitutes an outstanding event with respect to technical safety which can be used to describe the phenomenology of the early as well as the late phases of core degradation, for modeling and, above all, for validation of codes analyzing very severe accidents and, in this way, can serve to enhance the safety of plants in operation. After a brief introduction, the accident scenario is outlined. On this basis, the ATHLET-CD code is introduced, and the approach used in modeling the plant and the accident is described. Finally, the results of the simulation carried out with ATHLET-CD are summarized and evaluated. It is seen that the code is able, in principle, to describe the accident with good accuracy. However, further development with respect to modeling of the late phase is required. (orig.)

  8. Analysis of the TMI-2 accident using ATHLET-CD

    Drath, T.; Kleinhietpass, I.D.; Koch, M.K. [Lehrstuhl fuer Energiesysteme und Energiewirtschaft (LEE), Ruhr-Univ. Bochum (RUB) (Germany)

    2006-01-01

    On March 28, 1979, the only loss-of-coolant accident so far in a light water reactor exceeding the design basis occurred in unit 2 of the U.S. PWR of Three Mile Island (TMI-2) near Harrisburg. The accident entailed massive core degradation accompanied by the formation of a bed of debris and a molten pool. Approx. 30 t of the core inventory were moved to the bottom of the reactor pressure vessel; the vessel sustained this thermal load. Also because of the worldwide use of light water reactors, this accident constitutes an outstanding event with respect to technical safety which can be used to describe the phenomenology of the early as well as the late phases of core degradation, for modeling and, above all, for validation of codes analyzing very severe accidents and, in this way, can serve to enhance the safety of plants in operation. After a brief introduction, the accident scenario is outlined. On this basis, the ATHLET-CD code is introduced, and the approach used in modeling the plant and the accident is described. Finally, the results of the simulation carried out with ATHLET-CD are summarized and evaluated. It is seen that the code is able, in principle, to describe the accident with good accuracy. However, further development with respect to modeling of the late phase is required. (orig.)

  9. Estimation of doses received by operators in the 1958 RB reactor accident using the MCNP5 computer code simulation

    Pešić Milan P.

    2012-01-01

    Full Text Available A numerical simulation of the radiological consequences of the RB reactor reactivity excursion accident, which occurred on October 15, 1958, and an estimation of the total doses received by the operators were run by the MCNP5 computer code. The simulation was carried out under the same assumptions as those used in the 1960 IAEA-organized experimental simulation of the accident: total fission energy of 80 MJ released in the accident and the frozen positions of the operators. The time interval of exposure to high doses received by the operators has been estimated. Data on the RB1/1958 reactor core relevant to the accident are given. A short summary of the accident scenario has been updated. A 3-D model of the reactor room and the RB reactor tank, with all the details of the core, created. For dose determination, 3-D simplified, homogenised, sexless and faceless phantoms, placed inside the reactor room, have been developed. The code was run for a number of neutron histories which have given a dose rate uncertainty of less than 2%. For the determination of radiation spectra escaping the reactor core and radiation interaction in the tissue of the phantoms, the MCNP5 code was run (in the KCODE option and “mode n p e”, with a 55-group neutron spectra, 35-group gamma ray spectra and a 10-group electron spectra. The doses were determined by using the conversion of flux density (obtained by the F4 tally in the phantoms to doses using factors taken from ICRP-74 and from the deposited energy of neutrons and gamma rays (obtained by the F6 tally in the phantoms’ tissue. A rough estimation of the time moment when the odour of ozone was sensed by the operators is estimated for the first time and given in Appendix A.1. Calculated total absorbed and equivalent doses are compared to the previously reported ones and an attempt to understand and explain the reasons for the obtained differences has been made. A Root Cause Analysis of the accident was done and

  10. Loss of coolant accident analysis for the IEA-R1 reactor at 5 MW

    The postulated Loss of Coolant Accidents (LOCA) for the IEA-R1 Brazilian Research operating at 5 MW are qualitatively analyzed. Two computer codes, LOSS and TEMPLOCA, were developed. The computer code LOSS determines the time to drain the pool down to the level of the bottom of the core, and the computer code TEMPLOCA calculates the peak fuel element temperature during the transient. Four groups of accidents were analyzed: damaged pool, pump-down of pool, failure of beam tubes or other penetrations, and Primary coolant rupture. The analysis showed the necessity of introducing a new Emergency Core Cooling System (ECCS), and appropriate modifications in other systems and components of the reactor, such as: Primary Coolant System, Pneumatic Tube System and beam tubes. The effectiveness of the ECCS will ensure that the reactor can be safely operated at 5 MW and that can withstand a loss of coolant accident without sustaining core damage. (author)