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Sample records for accident development modeling

  1. Accident sequence precursor analysis level 2/3 model development

    International Nuclear Information System (INIS)

    Lui, C.H.; Galyean, W.J.; Brownson, D.A.

    1997-01-01

    The US Nuclear Regulatory Commission's Accident Sequence Precursor (ASP) program currently uses simple Level 1 models to assess the conditional core damage probability for operational events occurring in commercial nuclear power plants (NPP). Since not all accident sequences leading to core damage will result in the same radiological consequences, it is necessary to develop simple Level 2/3 models that can be used to analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude of the resulting radioactive releases to the environment, and calculate the consequences associated with these releases. The simple Level 2/3 model development work was initiated in 1995, and several prototype models have been completed. Once developed, these simple Level 2/3 models are linked to the simple Level 1 models to provide risk perspectives for operational events. This paper describes the methods implemented for the development of these simple Level 2/3 ASP models, and the linkage process to the existing Level 1 models

  2. Severe accident development modeling and evaluation for CANDU

    Energy Technology Data Exchange (ETDEWEB)

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

    2009-09-15

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

  3. Severe accident development modeling and evaluation for CANDU

    International Nuclear Information System (INIS)

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

    2009-01-01

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

  4. The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

    Directory of Open Access Journals (Sweden)

    Ludfi Pratiwi Bowo

    2017-06-01

    Full Text Available Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequential techniques, epidemiological techniques and systemic techniques, where the marine accidents are included in the epidemiological technique. This study compares the Human Error Assessment and Reduction Technique (HEART methodology and the 4M Overturned Pyramid (MOP model, which are applied to assess marine accidents. Furthermore, the MOP model can effectively describe the relationships of other factors which affect the accidents; whereas, the HEART methodology is only focused on human factors.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-04-01

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

  6. Development of A Methodology for Assessing Various Accident Management Strategies Using Decision Tree Models

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Nam Yeong; Kim, Jin Tae; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

    2016-05-15

    The purpose of ASP (Accident Sequence Precursor) analysis is to evaluate operational accidents in full power and low power operation by using PRA (Probabilistic Risk Assessment) technologies. The awareness of the importance of ASP analysis has been on rise. The methodology for ASP analysis has been developed in Korea, KINS (Korea Institute of Nuclear Safety) has managed KINS-ASP program since it was developed. In this study, we applied ASP analysis into operational accidents in full power and low power operation to quantify CCDP (Conditional Core Damage Probability). To reflect these 2 cases into PRA model, we modified fault trees and event trees of the existing PRA model. Also, we suggest the ASP regulatory system in the conclusion. In this study, we reviewed previous studies for ASP analysis. Based on it, we applied it into operational accidents in full power and low power operation. CCDP of these 2 cases are 1.195E-06 and 2.261E-03. Unlike other countries, there is no regulatory basis of ASP analysis in Korea. ASP analysis could detect the risk by assessing the existing operational accidents. ASP analysis can improve the safety of nuclear power plant by detecting, reviewing the operational accidents, and finally removing potential risk. Operator have to notify regulatory institute of operational accident before operator takes recovery work for the accident. After follow-up accident, they have to check precursors in data base to find similar accident.

  7. Contribution to Accident Prediction Models Development for Rural Two-Lane Roads in Serbia

    Directory of Open Access Journals (Sweden)

    Draženko Glavić

    2016-08-01

    Full Text Available Over the last three decades numerous research efforts have been conducted worldwide to determine the relationship between traffic accidents and traffic and road characteristics. So far, the mentioned studies have not been carried out in Serbia and in the region. This paper represents one of the first attempts to develop accident prediction models in Serbia. The paper provides a comprehensive literature review, describes procedures for collection and analysis of the traffic accident data, as well as the methodology used to develop the accident prediction models. The paper presents models obtained by both univariate and multivariate regression analyses. The obtained results are compared to the results of other studies and comparisons are discussed. Finally, the paper presents conclusions and important points for future research. The results of this research can find theoretical as well as practical application.

  8. Development on Dose Assessment Model of Northeast Asia Nuclear Accident Simulator

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Ju Yub; Kim, Ju Youl; Kim, Suk Hoon; Lee, Seung Hee; Yoon, Tae Bin [FNC Techology, Yongin (Korea, Republic of)

    2016-05-15

    In order to support the emergency response system, the simulator for overseas nuclear accident is under development including source-term estimation, atmospheric dispersion modeling and dose assessment. The simulator is named NANAS (Northeast Asia Nuclear Accident Simulator). For the source-term estimation, design characteristics of each reactor type should be reflected into the model. Since there are a lot of reactor types in neighboring countries, the representative reactors of China, Japan and Taiwan have been selected and the source-term estimation models for each reactor have been developed, respectively. For the atmospheric dispersion modeling, Lagrangian particle model will be integrated into the simulator for the long range dispersion modeling in Northeast Asia region. In this study, the dose assessment model has been developed considering external and internal exposure. The dose assessment model has been developed as a part of the overseas nuclear accidents simulator which is named NANAS. It addresses external and internal pathways including cloudshine, groundshine and inhalation. Also, it uses the output of atmospheric dispersion model (i.e. the average concentrations of radionuclides in air and ground) and various coefficients (e.g. dose conversion factor and breathing rate) as an input. Effective dose and thyroid dose for each grid in the Korean Peninsula region are printed out as a format of map projection and chart. Verification and validation on the dose assessment model will be conducted in further study by benchmarking with the measured data of Fukushima Daiichi Nuclear Accident.

  9. Development of the model for the stress calculation of fuel assembly under accident load

    International Nuclear Information System (INIS)

    Kim, Il Kon

    1993-01-01

    The finite element model for the stress calculation in guide thimbles of a fuel assembly (FA) under seismic and loss-of-coolant-accident (LOCA) load is developed. For the stress calculation of FA under accident load, at first the program MAIN is developed to select the worst bending mode shaped FA from core model. And then the model for the stress calculation of FA is developed by means of the finite element code. The calculated results of program MAIN are used as the kinematic constraints of the finite element model of a FA. Compared the calculated results of the stiffness of the finite element model of FA with the test results they have good agreements. (Author)

  10. German offsite accident consequence model for nuclear facilities: further development and application

    International Nuclear Information System (INIS)

    Bayer, A.

    1985-01-01

    The German Offsite Accident Consequence Model - first applied in the German Risk Study for nuclear power plants with light water reactors - has been further developed with the improvement of several important submodels in the areas of atmospheric dispersion, shielding effects of houses, and the foodchains. To aid interpretation, the presentation of results has been extended with special emphasis on the presentation of the loss of life expectancy. The accident consequence model has been further developed for application to risk assessments for other nuclear facilities, e.g., the liquid metal fast breeder reactor (SNR-300) and the high temperature gas cooled reactor. Moreover the model have been further developed in the area of optimal countermeasure strategies (sheltering, evacuation, etc.) in the case of the Central European conditions. Preliminary considerations has been performed in connection with safety goals on the basis of doses

  11. Developments in modelling the economic impact of Off-site accident consequences

    International Nuclear Information System (INIS)

    Haywood, S.M.; Robinson, C.A.; Faude, D.

    1991-01-01

    Models for assessing the economic consequences of accidental releases of radioactivity have application both in accident consequence codes and in decision aiding computer systems for use in emergency response. Such models may be applied in emergency planning, and studies in connection with the siting, design and licensing of nuclear facilities. Several models for predicting economic impact have been developed, in Europe and the US, and these are reviewed. A new model, called COCO-1 (Cost of Consequences Off-site), has been developed under the CEC MARIA programme and the features of the model are summarised. The costs calculated are a measure of the benefit foregone as a result of the accident, and in addition to tangible monetary costs the model attempts to include costs arising from the effect of the accident on individuals, for instance the disruption caused by the loss of homes. COCO-1 includes the cost of countermeasures, namely evacuation, relocation, sheltering, food restrictions and decontamination, and also the cost of health effects in the exposed population. The primary quantity used in COCO-1 to measure the economic value of land subject to restrictions on usage is Gross Domestic Product (GDP). Examples of default data included in the model are presented, as are the results of an illustrative application. The limitations of COCO-1 are discussed, and areas where further data are needed are identified

  12. The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

    OpenAIRE

    Ludfi Pratiwi Bowo; Wanginingastuti Mutmainnah; Masao Furusho

    2017-01-01

    Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA) that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequen...

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2017-12-23

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

  14. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    Homma, T.; Togawa, O.

    1991-01-01

    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)

  15. The development and demonstration of integrated models for the evaluation of severe accident management strategies - SAMEM

    International Nuclear Information System (INIS)

    Ang, M.L.; Peers, K.; Kersting, E.; Fassmann, W.; Tuomisto, H.; Lundstroem, P.; Helle, M.; Gustavsson, V.; Jacobsson, P.

    2001-01-01

    This study is concerned with the further development of integrated models for the assessment of existing and potential severe accident management (SAM) measures. This paper provides a brief summary of these models, based on Probabilistic Safety Assessment (PSA) methods and the Risk Oriented Accident Analysis Methodology (ROAAM) approach, and their application to a number of case studies spanning both preventive and mitigative accident management regimes. In the course of this study it became evident that the starting point to guide the selection of methodology and any further improvement is the intended application. Accordingly, such features as the type and area of application and the confidence requirement are addressed in this project. The application of an integrated ROAAM approach led to the implementation, at the Loviisa NPP, of a hydrogen mitigation strategy, which requires substantial plant modifications. A revised level 2 PSA model was applied to the Sizewell B NPP to assess the feasibility of the in-vessel retention strategy. Similarly the application of PSA based models was extended to the Barseback and Ringhals 2 NPPs to improve the emergency operating procedures, notably actions related to manual operations. A human reliability analysis based on the Human Cognitive Reliability (HCR) and Technique For Human Error Rate (THERP) models was applied to a case study addressing secondary and primary bleed and feed procedures. Some aspects pertinent to the quantification of severe accident phenomena were further examined in this project. A comparison of the applications of PSA based approach and ROAAM to two severe accident issues, viz hydrogen combustion and in-vessel retention, was made. A general conclusion is that there is no requirement for further major development of the PSA and ROAAM methodologies in the modelling of SAM strategies for a variety of applications as far as the technical aspects are concerned. As is demonstrated in this project, the

  16. Development of hydrogeological modelling approaches for assessment of consequences of hazardous accidents at nuclear power plants

    International Nuclear Information System (INIS)

    Rumynin, V.G.; Mironenko, V.A.; Konosavsky, P.K.; Pereverzeva, S.A.

    1994-07-01

    This paper introduces some modeling approaches for predicting the influence of hazardous accidents at nuclear reactors on groundwater quality. Possible pathways for radioactive releases from nuclear power plants were considered to conceptualize boundary conditions for solving the subsurface radionuclides transport problems. Some approaches to incorporate physical-and-chemical interactions into transport simulators have been developed. The hydrogeological forecasts were based on numerical and semi-analytical scale-dependent models. They have been applied to assess the possible impact of the nuclear power plants designed in Russia on groundwater reservoirs

  17. Analysis of multiple failure accident scenarios for development of probabilistic safety assessment model for KALIMER-600

    International Nuclear Information System (INIS)

    Kim, T.W.; Suk, S.D.; Chang, W.P.; Kwon, Y.M.; Jeong, H.Y.; Lee, Y.B.; Ha, K.S.; Kim, S.J.

    2009-01-01

    A sodium-cooled fast reactor (SFR), KALIMER-600, is under development at KAERI. Its fuel is the metal fuel of U-TRU-Zr and it uses sodium as coolant. Its advantages are found in the aspects of an excellent uranium resource utilization, inherent safety features, and nonproliferation. The probabilistic safety assessment (PSA) will be one of the initiating subjects for designing it from the aspects of a risk informed design (RID) as well as a technology-neutral licensing (TNL). The core damage is defined as coolant voiding, fuel melting, or cladding damage. Accident scenarios which lead to the core damage should be identified for the development of a Level-1 PSA model. The SSC-K computer code is used to identify the conditions which lead to core damage. KALIMER-600 has passive safety features such as passive shutdown functions, passive pump coast-down features, and passive decay heat removal systems. It has inherent reactivity feedback effects such as Doppler, sodium void, core axial expansion, control rod axial expansion, core radial expansion, etc. The accidents which are analyzed are the multiple failure accidents such as an unprotected transient overpower, a loss of flow, and a loss of heat sink events with degraded safety systems or functions. The safety functions to be considered here are a reactor trip, inherent reactivity feedback features, the pump coast-down, and the passive decay heat removal. (author)

  18. Development of CHF models for inner and outer RPV gaps in a meltdown severe accident

    International Nuclear Information System (INIS)

    Wang, J.; Tian, W.X.; Feng, K.; Yu, H.X.; Zhang, Y.P.; Su, G.H.; Qiu, S.Z.

    2013-01-01

    Highlights: • A CHF model was developed to predict the CHF in hemispherical narrow gap. • The computed result was validated by the test data of Park and Köhler. • An analytical CHF model was developed to predict the CHF on the outer surface of the lower head. • The predicted CHF was compared with the experimental data of ULPU-V. • Two CHF models developed for the inner and outer CHF predict the CHF well. - Abstract: During a severe accident, the core melt relocates in the lower head and a hemispherical narrow gap may appear between the crust and the lower head because of the different material expansion ratio. The existence of this gap is very important to the integrity of the lower head. Based on the counter current flow limitation (CCFL) between the vapor phase and the liquid phase, a CHF model was developed to predict the CHF in hemispherical narrow gap. The CHF model developed was validated by the test data of Park and Köhler. The effect of key parameters, including the system pressure, radius of melt, and gap size, on the CHF were investigated. And the TMI-2 accident was also calculated by using the CHF formula. Moreover, based on the interface separation model, an analytical CHF model was developed to predict the CHF on the outer surface of the lower head. The predicted CHF was compared with the experimental data of ULPU-V. It indicated that the CHF models developed for the inner and outer CHF could predict the CHF well

  19. Development of systematic models for aerosol agglomeration and spray removal under severe accident conditions

    International Nuclear Information System (INIS)

    Kajimoto, Mitsuhiro

    2008-01-01

    Radionuclide behavior during various severe accident conditions has been addressed as one of the important issues to discuss environmental safety in nuclear power plants. The present paper deals with the development of analytical models and their validations for the agglomeration of multiple-component aerosol and spray removal that controls source terms to the environment of both aerosols and gaseous radionuclides during recirculation mode operation in a containment system for a light water reactor. As for aerosol agglomeration, the single collision kernel model that can cover all types of two-body collision of aerosol was developed. In addition, the dynamic model that can treat aerosol and vapor transfer leading to the equilibrium condition under the containment spray operation was developed. The validations of the present models for multiple-component aerosol growth by agglomeration were performed by comparisons with Nuclear Safety Pilot Plant (NSPP) experiments at Oak Ridge National Laboratory (ORNL) and AB experiments at Hanford Engineering National Laboratory (HEDL). In addition, the spray removal models were applied to the analysis of containment spray experiment (CSE) at HEDL. The results calculated by the models showed good agreements with experimental results. (author)

  20. Development of a PWR-W GOTHIC 3D model for containment accident analysis

    International Nuclear Information System (INIS)

    Bocanegra, Rafael; Jimenez, Gonzalo; Fernández-Cosials, Mikel Kevin

    2016-01-01

    Highlights: • The development of several 3D PWR containment models is described. • A Large Break LOCA is simulated. • The temperature and velocity fields are highly dependent on three-dimensional phenomena. • The pressure evolution is qualitatively similar in all models with small quantitative differences. - Abstract: The confinement of radioactive material in a nuclear power plant, including the discharge control and the release minimization, is a fundamental safety function to be ensured in a design basis accident (DBA). For plant licensing analysis, the containment is usually modeled with a lumped parameter approach. Inherent to the lumped parameter approach is the assumption that within each region the fluid is well mixed. However, the containment is a large building with a complex configuration and it is distributed in several compartments that avoid the well mixing of the fluid and could have three-dimensional effects that affect the thermal–hydraulic behavior. Therefore, the commonly used lumped parameter approach may not be enough to capture these effects. In order to study these assumptions, four generic PWR containment models have been developed for Mass and Energy (M&E) release analysis with GOTHIC 8.0 (QA) code, three of them being subdivided and the fourth one is a lumped parameter model. A Large Break LOCA is simulated in order to compare the thermal–hydraulic behavior of the different models. The results show a high dependence on the three-dimensional phenomena, especially the temperature and velocity distribution. In contrast, the pressure evolution is qualitatively similar in all models with small quantitative differences.

  1. Modelling Accident Tolerant Fuel Concepts

    Energy Technology Data Exchange (ETDEWEB)

    Hales, Jason Dean [Idaho National Laboratory; Gamble, Kyle Allan Lawrence [Idaho National Laboratory

    2016-05-01

    The catastrophic events that occurred at the Fukushima-Daiichi nuclear power plant in 2011 have led to widespread interest in research of alternative fuels and claddings that are proposed to be accident tolerant. The United States Department of Energy (DOE) through its Nuclear Energy Advanced Modeling and Simulation (NEAMS) program has funded an Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The ATF HIP is a three-year project to perform research on two accident tolerant concepts. The final outcome of the ATF HIP will be an in-depth report to the DOE Advanced Fuels Campaign (AFC) giving a recommendation on whether either of the two concepts should be included in their lead test assembly scheduled for placement into a commercial reactor in 2022. The two ATF concepts under investigation in the HIP are uranium silicide fuel and iron-chromium-aluminum (FeCrAl) alloy cladding. Utilizing the expertise of three national laboratory participants (Idaho National Laboratory, Los Alamos National Laboratory, and Argonne National Laboratory), a comprehensive multiscale approach to modeling is being used that includes atomistic modeling, molecular dynamics, rate theory, phase-field, and fuel performance simulations. Model development and fuel performance analysis are critical since a full suite of experimental studies will not be complete before AFC must prioritize concepts for focused development. In this paper, we present simulations of the two proposed accident tolerance fuel systems: U3Si2 fuel with Zircaloy-4 cladding, and UO2 fuel with FeCrAl cladding. Sensitivity analyses are completed using Sandia National Laboratories’ Dakota software to determine which input parameters (e.g., fuel specific heat) have the greatest influence on the output metrics of interest (e.g., fuel centerline temperature). We also outline the multiscale modelling approach being employed. Considerable additional work is required prior to preparing the recommendation report for the Advanced

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

    Energy Technology Data Exchange (ETDEWEB)

    Saghafi, Mahdi [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); Ghofrani, Mohammad Bagher, E-mail: ghofrani@sharif.edu [Department of Energy Engineering, Sharif University of Technology, Azadi Avenue, Tehran (Iran, Islamic Republic of); D’Auria, Francesco [San Piero a Grado Nuclear Research Group (GRNSPG), University of Pisa, Via Livornese 1291, San Piero a Grado, Pisa (Italy)

    2016-07-15

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

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

  4. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed.

  5. Development of two-dimensional hot pool model and analysis of the ULOHS accident in KALIMER design

    International Nuclear Information System (INIS)

    Lee, Yong Bum; Jeong, K. S.; Hahn, H. D.

    2000-10-01

    In the new version of HP2D program, the variation model of the hot pool sodium level is added so that the temperature and velocity profiles can be predicted more accurately than old version. To verify and validate the developed new version model, comparison of the MONJU experimental data with the predicted one is performed and analyzed. And also the ULOHS(Unprotected Loss of Heat Sink) accident in the KALIMER design is performed and analyzed

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

    International Nuclear Information System (INIS)

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

    2002-01-01

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

  7. Development of simplified 1D and 2D models for studying a PWR lower head failure under severe accident conditions

    International Nuclear Information System (INIS)

    Koundy, V.; Dupas, J.; Bonneville, H.; Cormeau, I.

    2005-01-01

    In the study of severe accidents of nuclear pressurized water reactors, the scenarios that describe the relocation of significant quantities of liquid corium at the bottom of the lower head are investigated from the mechanical point of view. In these scenarios, the risk of a breach and the possibility of a large quantity of corium being released from the lower head exist. This may lead to direct heating of the containment or outer vessel steam explosion. These issues are important due to their early containment failure potential. Since the TMI-2 accident, many theoretical and experimental investigations, relating to lower head mechanical behaviour under severe thermo-mechanical loading in the event of a core meltdown accident have been performed. IRSN participated actively in the one-fifth scale USNRC/SNL LHF and OECD LHF (OLHF) programs. Within the framework of these programs, two simplified models were developed by IRSN: the first is a simplified 1D approach based on the theory of pressurized spherical shells and the second is a simplified 2D model based on the theory of shells of revolution under symmetric loading. The mathematical formulation of both models and the creep constitutive equations used are presented in detail in this paper. The corresponding models were used to interpret some of the OLHF program experiments and the calculation results were quite consistent with the experimental data. The two simplified models have been used to simulate the thermo-mechanical behaviour of a 900 MWe pressurized water reactor lower head under severe accident conditions leading to failure. The average transient heat flux produced by the corium relocated at the bottom of the lower head has been determined using the IRSN HARAR code. Two different methods, both taking into account the ablation of the internal surface, are used to determine the temperature profiles across the lower head wall and their effect on the time to failure is discussed. Using these simplified models

  8. Modeling accidents for prioritizing prevention

    International Nuclear Information System (INIS)

    Hale, A.R.; Ale, B.J.M.; Goossens, L.H.J.; Heijer, T.; Bellamy, L.J; Mud, M.L.; Roelen, A.; Baksteen, H.; Post, J.; Papazoglou, I.A.; Bloemhoff, A.; Oh, J.I.H.

    2007-01-01

    The Workgroup Occupational Risk Model (WORM) project in the Netherlands is developing a comprehensive set of scenarios to cover the full range of occupational accidents. The objective is to support companies in their risk analysis and prioritization of prevention. This paper describes how the modeling has developed through projects in the chemical industry, to this one in general industry and how this is planned to develop further in the future to model risk prevention in air transport. The core modeling technique is based on the bowtie, with addition of more explicit modeling of the barriers needed for risk control, the tasks needed to ensure provision, use, monitoring and maintenance of the barriers, and the management resources and tasks required to ensure that these barrier life cycle tasks are carried out effectively. The modeling is moving from a static notion of barriers which can fail, to seeing risk control dynamically as (fallible) means for staying within a safe envelope. The paper shows how concepts develop slowly over a series of projects as a core team works continuously together. It concludes with some results of the WORM project and some indications of how the modeling is raising fundamental questions about the conceptualization of system safety, which need future resolution

  9. Adverse fetal outcome in road accidents: Injury mechanism study and injury criteria development in a pregnant woman finite element model.

    Science.gov (United States)

    Auriault, F; Thollon, L; Pérès, J; Behr, M

    2016-12-01

    This study documents the development of adverse fetal outcome predictors dedicated to the analysis of road accidents involving pregnant women. To do so, a pre-existing whole body finite element model representative of a 50th percentile 26 weeks pregnant woman was used. A total of 8 accident scenarios were simulated with the model positioned on a sled. Each of these scenarios was associated to a risk of adverse fetal outcome based on results from real car crash investigations involving pregnant women from the literature. The use of airbags and accidents involving unbelted occupants were not considered in this study. Several adverse fetal outcome potential predictors were then evaluated with regard to their correlation to this risk of fetal injuries. Three predictors appeared strongly correlated to the risk of adverse fetal outcome: (1) the intra uterine pressure at the placenta fetal side area (r=0.92), (2) the fetal head acceleration (HIC) (r=0.99) and (3) area of utero-placental interface over a strain threshold (r=0.90). Finally, sensitivity analysis against slight variations of the simulation parameters was performed and assess robustness of these criteria. Copyright © 2016 Elsevier Ltd. All rights reserved.

  10. Key Characteristics of Combined Accident including TLOFW accident for PSA Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bo Gyung; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Technology and Research, Abu Dhabi (United Arab Emirates)

    2015-05-15

    accident is very hazardous. If the operator initiates F and B operation properly under the combined accident including TLOFW accident, the operators can prevent the core damage. Since F and B operation is last resort to prevent core damage and necessary conditions of F and B operation are very complicated, the consequence of these events should be considered in PSA model to improve emergency response capabilities under the rare events. Dynamic PSA modeling is better to estimate the effects of heading order and timing issues. Especially, dynamic PSA can model accident sequences and estimate their probabilities through integrated, time-dependent, probabilistic and deterministic models of NPPs, based on the thermal-hydraulic processes and operator behavior in accident conditions. We will develop the dynamic PSA model for the combined accident including TLOFW accident in the further study.

  11. Major Accidents (Gray Swans) Likelihood Modeling Using Accident Precursors and Approximate Reasoning.

    Science.gov (United States)

    Khakzad, Nima; Khan, Faisal; Amyotte, Paul

    2015-07-01

    Compared to the remarkable progress in risk analysis of normal accidents, the risk analysis of major accidents has not been so well-established, partly due to the complexity of such accidents and partly due to low probabilities involved. The issue of low probabilities normally arises from the scarcity of major accidents' relevant data since such accidents are few and far between. In this work, knowing that major accidents are frequently preceded by accident precursors, a novel precursor-based methodology has been developed for likelihood modeling of major accidents in critical infrastructures based on a unique combination of accident precursor data, information theory, and approximate reasoning. For this purpose, we have introduced an innovative application of information analysis to identify the most informative near accident of a major accident. The observed data of the near accident were then used to establish predictive scenarios to foresee the occurrence of the major accident. We verified the methodology using offshore blowouts in the Gulf of Mexico, and then demonstrated its application to dam breaches in the United Sates. © 2015 Society for Risk Analysis.

  12. Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

    Energy Technology Data Exchange (ETDEWEB)

    Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho [KAERI, Deajeon (Korea, Republic of)

    2016-05-15

    The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated.

  13. Developing Off-site Emergency Preparedness and Response Model (OEPRM) for Severe Accident of NPP in a Densely Populated Country Using System Dynamics Approach

    International Nuclear Information System (INIS)

    Hossena, Muhammed Mufazzal; Kang, Kyoung Ho; Song, Jin Ho

    2016-01-01

    The main objectives of this study are to find the influencing factors of systems and sub-systems of OEPRM, to find the interdependency among the influencing factors, and to develop a conceptual qualitative OEPRM for densely populated NPP country in case of SA using system dynamics (SD). NPP accidents are classified as nuclear accidents and incidents depending on the severity. Severe accident (SA) is certain low probability accident that are beyond design basis accident which may arise due to multiple failures of safety systems leading to significant core degradation and jeopardize the integrity of many or all of the barriers to the release of radioactive material. The weakness to the off-site emergency response in the time of Fukushima accident was observed. So, it is crucial to develop an off-site emergency preparedness and responses model (OEPRM) for radiological emergency in densely populated country from the Fukushima emergency response lesson. In this study, an OEPRM is developed for densely populated NPP country to mitigate radiological effects in case of SA using SD approach. Besides, this study focuses the weakness of emergency response in Fukushima accident and proposed solution approach. The development of OEPRM in case of SA of NPP is very complex because of the involvement of various organization and it requires highly specialized agencies and technical experts. Moreover, if the country is agriculture based, it will make completely sophisticated

  14. Developing a Model for Forecasting Road Traffic Accident (RTA Fatalities in Yemen

    Directory of Open Access Journals (Sweden)

    Karim Fareed M. A.

    2017-12-01

    Full Text Available The aim of this paper is to develop a model for forecasting RTA fatalities in Yemen. The yearly fatalities was modeled as the dependent variable, while the number of independent variables included the population, number of vehicles, GNP, GDP and Real GDP per capita. It was determined that all these variables are highly correlated with the correlation coefficient (r ≈ 0.9; in order to avoid multicollinearity in the model, a single variable with the highest r value was selected (real GDP per capita. A simple regression model was developed; the model was very good (R2=0.916; however, the residuals were serially correlated. The Prais-Winsten procedure was used to overcome this violation of the regression assumption. The data for a 20-year period from 1991-2010 were analyzed to build the model; the model was validated by using data for the years 2011-2013; the historical fit for the period 1991 - 2011 was very good. Also, the validation for 2011-2013 proved accurate.

  15. Developing a Model for Forecasting Road Traffic Accident (RTA) Fatalities in Yemen

    Science.gov (United States)

    Karim, Fareed M. A.; Abdo Saleh, Ali; Taijoobux, Aref; Ševrović, Marko

    2017-12-01

    The aim of this paper is to develop a model for forecasting RTA fatalities in Yemen. The yearly fatalities was modeled as the dependent variable, while the number of independent variables included the population, number of vehicles, GNP, GDP and Real GDP per capita. It was determined that all these variables are highly correlated with the correlation coefficient (r ≈ 0.9); in order to avoid multicollinearity in the model, a single variable with the highest r value was selected (real GDP per capita). A simple regression model was developed; the model was very good (R2=0.916); however, the residuals were serially correlated. The Prais-Winsten procedure was used to overcome this violation of the regression assumption. The data for a 20-year period from 1991-2010 were analyzed to build the model; the model was validated by using data for the years 2011-2013; the historical fit for the period 1991 - 2011 was very good. Also, the validation for 2011-2013 proved accurate.

  16. Development of zircaloy deformation model to describe the zircaloy-4 cladding tube during accidents

    International Nuclear Information System (INIS)

    Raff, S.

    1978-01-01

    The development of a high-temperature deformation model for Zircaloy-4 cans is primarily based on numerous well-parametrized tensile tests to get the material behaviour including statistical variance. It is shown that plastic deformation may be described by a power creep law, the coefficients of which show strong dependence on temperature in the relevant temperature region. These coefficients have been determined. A model based on these coefficients has been established which, apart from best estimate deformation, gives upper and lower bounds of possible deformation. The model derived from isothermal uniaxial tests is being verified against isothermal and transient tube burst tests. The influence of preoxidation and increased oxygen concentration during deformation is modeled on the basis of the pseudobinary Zircaloy-oxygen phase diagram. (author)

  17. Development of a deformation and failure model for Zircaloy at high temperatures for light water reactor loss-of-coolant-accident investigations

    International Nuclear Information System (INIS)

    Raff, S.

    1982-11-01

    To describe Zircaloy-4 deformation and failure behaviour at high temperatures (600 to 1400 0 C), the phenomenological model NORA was developed and verified against numerous experimental results. The model can be applied to the calculation of fuel rod cladding deformation during small and large break loss-of-coolant-accidents. (orig./RW) [de

  18. A study on industrial accident rate forecasting and program development of estimated zero accident time in Korea.

    Science.gov (United States)

    Kim, Tae-gu; Kang, Young-sig; Lee, Hyung-won

    2011-01-01

    To begin a zero accident campaign for industry, the first thing is to estimate the industrial accident rate and the zero accident time systematically. This paper considers the social and technical change of the business environment after beginning the zero accident campaign through quantitative time series analysis methods. These methods include sum of squared errors (SSE), regression analysis method (RAM), exponential smoothing method (ESM), double exponential smoothing method (DESM), auto-regressive integrated moving average (ARIMA) model, and the proposed analytic function method (AFM). The program is developed to estimate the accident rate, zero accident time and achievement probability of an efficient industrial environment. In this paper, MFC (Microsoft Foundation Class) software of Visual Studio 2008 was used to develop a zero accident program. The results of this paper will provide major information for industrial accident prevention and be an important part of stimulating the zero accident campaign within all industrial environments.

  19. Modeling secondary accidents identified by traffic shock waves.

    Science.gov (United States)

    Junhua, Wang; Boya, Liu; Lanfang, Zhang; Ragland, David R

    2016-02-01

    The high potential for occurrence and the negative consequences of secondary accidents make them an issue of great concern affecting freeway safety. Using accident records from a three-year period together with California interstate freeway loop data, a dynamic method for more accurate classification based on the traffic shock wave detecting method was used to identify secondary accidents. Spatio-temporal gaps between the primary and secondary accident were proven be fit via a mixture of Weibull and normal distribution. A logistic regression model was developed to investigate major factors contributing to secondary accident occurrence. Traffic shock wave speed and volume at the occurrence of a primary accident were explicitly considered in the model, as a secondary accident is defined as an accident that occurs within the spatio-temporal impact scope of the primary accident. Results show that the shock waves originating in the wake of a primary accident have a more significant impact on the likelihood of a secondary accident occurrence than the effects of traffic volume. Primary accidents with long durations can significantly increase the possibility of secondary accidents. Unsafe speed and weather are other factors contributing to secondary crash occurrence. It is strongly suggested that when police or rescue personnel arrive at the scene of an accident, they should not suddenly block, decrease, or unblock the traffic flow, but instead endeavor to control traffic in a smooth and controlled manner. Also it is important to reduce accident processing time to reduce the risk of secondary accident. Copyright © 2015 Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  1. Using Numerical Models in the Development of Software Tools for Risk Management of Accidents with Oil and Inert Spills

    Science.gov (United States)

    Fernandes, R.; Leitão, P. C.; Braunschweig, F.; Lourenço, F.; Galvão, P.; Neves, R.

    2012-04-01

    The increasing ship traffic and maritime transport of dangerous substances make it more difficult to significantly reduce the environmental, economic and social risks posed by potential spills, although the security rules are becoming more restrictive (ships with double hull, etc.) and the surveillance systems are becoming more developed (VTS, AIS). In fact, the problematic associated to spills is and will always be a main topic: spill events are continuously happening, most of them unknown for the general public because of their small scale impact, but with some of them (in a much smaller number) becoming authentic media phenomena in this information era, due to their large dimensions and environmental and social-economic impacts on ecosystems and local communities, and also due to some spectacular or shocking pictures generated. Hence, the adverse consequences posed by these type of accidents, increase the preoccupation of avoiding them in the future, or minimize their impacts, using not only surveillance and monitoring tools, but also increasing the capacity to predict the fate and behaviour of bodies, objects, or substances in the following hours after the accident - numerical models can have now a leading role in operational oceanography applied to safety and pollution response in the ocean because of their predictive potential. Search and rescue operation, oil, inert (ship debris, or floating containers), and HNS (hazardous and noxious substances) spills risk analysis are the main areas where models can be used. Model applications have been widely used in emergency or planning issues associated to pollution risks, and contingency and mitigation measures. Before a spill, in the planning stage, modelling simulations are used in environmental impact studies, or risk maps, using historical data, reference situations, and typical scenarios. After a spill, the use of fast and simple modelling applications allow to understand the fate and behaviour of the spilt

  2. Methodological guidelines for developing accident modification functions

    DEFF Research Database (Denmark)

    Elvik, Rune

    2015-01-01

    This paper proposes methodological guidelines for developing accident modification functions. An accident modification function is a mathematical function describing systematic variation in the effects of road safety measures. The paper describes ten guidelines. An example is given of how to use...... limitations in developing accident modification functions are the small number of good evaluation studies and the often huge variation in estimates of effect. It is therefore still not possible to develop accident modification functions for very many road safety measures. © 2015 Elsevier Ltd. All rights...... the guidelines. The importance of exploratory analysis and an iterative approach in developing accident modification functions is stressed. The example shows that strict compliance with all the guidelines may be difficult, but represents a level of stringency that should be strived for. Currently the main...

  3. Correspondence model of occupational accidents

    Directory of Open Access Journals (Sweden)

    Juan C. Conte

    2011-09-01

    Full Text Available We present a new generalized model for the diagnosis and prediction of accidents among the Spanish workforce. Based on observational data of the accident rate in all Spanish companies over eleven years (7,519,732 accidents, we classified them in a new risk-injury contingency table (19×19. Through correspondence analysis, we obtained a structure composed of three axes whose combination identifies three separate risk and injury groups, which we used as a general Spanish pattern. The most likely or frequent relationships between the risk and injuries identified in the pattern facilitated the decision-making process in companies at an early stage of risk assessment. Each risk-injury group has its own characteristics, which are understandable within the phenomenological framework of the accident. The main advantages of this model are its potential application to any other country and the feasibility of contrasting different country results. One limiting factor, however, is the need to set a common classification framework for risks and injuries to enhance comparison, a framework that does not exist today. The model aims to manage work-related accidents automatically at any level.Apresentamos aqui um modelo generalizado para o diagnóstico e predição de acidentes na classe de trabalhadores da Espanha. Baseados em dados sobre a frequência de acidentes em todas as companhias da Espanha em 11 anos (7.519.732 acidentes, nós os classificamos em uma nova tabela de contingência risco-injúria (19×19. Através de uma análise por correspondência obtivemos uma estrutura composta por 3 eixos cuja combinação identifica 3 grupos separados de risco e injúria, que nós usamos como um perfil geral na Espanha. As mais prováveis ou frequentes relações entre risco e injúrias identificadas nesse perfil facilitaram o processo de decisão nas companhias em um estágio inicial de apreciação do risco. Cada grupo de risco-injúria tem suas próprias caracter

  4. Consequences of the Chernobyl accident as the basis for development of ecological and dosimetric models

    International Nuclear Information System (INIS)

    Balonov, M.I.

    1993-01-01

    The paper presents the materials of six year dynamic investigation of Cs and Sr radionuclides content in food products and body of inhabitants of contaminated regions of Russia, and of individual external doses of gamma radiation measured within thermoluminescent method. The models of formation of the dose of external gamma radiation and intake of Cs-134, -137 in body of inhabitants of contaminated locality are formed. The transfer factor from soil to vegetation and to food ration of man for Cs-134, 137 decreases with the half-period of 1 to 1.5 years and depends on soil type and agrochemical characteristics. (orig.)

  5. Development of integrated accident management assessment technology

    International Nuclear Information System (INIS)

    Jung, Won Dea; Ha, Jae Joo; Jin, Young Ho

    2002-04-01

    This project aims to develop critical technologies for accident management through securing evaluation frameworks and supporting tools, in order to enhance capabilities coping with severe accidents. For the research goal, firstly under the viewpoint of accident prevention, on-line risk monitoring system and the analysis framework for human error have been developed. Secondly, the training/supporting systems including the training simulator and the off-site risk evaluation system have been developed to enhance capabilities coping with severe accidents. Four kinds of research results have been obtained from this project. Firstly, the framework and taxonomy for human error analysis has been developed for accident management. As the second, the supporting system for accident managements has been developed. Using data that are obtained through the evaluation of off-site risk for Younggwang site, the risk database as well as the methodology for optimizing emergency responses has been constructed. As the third, a training support system, SAMAT, has been developed, which can be used as a training simulator for severe accident management. Finally, on-line risk monitoring system, DynaRM, has been developed for Ulchin 3 and 4 unit

  6. Model development of UO_2-Zr dispersion plate-type fuel behavior at early phase of severe accident and molten fuel meat relocation

    International Nuclear Information System (INIS)

    Zhang Zhuohua; Yu Junchong; Peng Shinian

    2014-01-01

    According to former study on oxygen diffusion, Nb-Zr solid reaction and UO_2-Zr solid reaction, the models of oxidation, solid reaction in fuel meat and relocation of molten fuel meat are developed based on structure and material properties of UO_2-Zr dispersion plate-type fuel, The new models can supply theoretical elements for the safety analysis of the core assembled with dispersion plate-type fuel under severe accident. (authors)

  7. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    International Nuclear Information System (INIS)

    Zhang, Yongfeng

    2016-01-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  8. Overview of lower length scale model development for accident tolerant fuels regarding U3Si2 fuel and FeCrAl cladding

    Energy Technology Data Exchange (ETDEWEB)

    Zhang, Yongfeng [Idaho National Laboratory

    2016-09-01

    U3Si2 and FeCrAl have been proposed as fuel and cladding concepts, respectively, for accident tolerance fuels with higher tolerance to accident scenarios compared to UO2. However, a lot of key physics and material properties regarding their in-pile performance are yet to be explored. To accelerate the understanding and reduce the cost of experimental studies, multiscale modeling and simulation are used to develop physics-based materials models to assist engineering scale fuel performance modeling. In this report, the lower-length-scale efforts in method and material model development supported by the Accident Tolerance Fuel (ATF) high-impact-problem (HIP) under the NEAMS program are summarized. Significant progresses have been made regarding interatomic potential, phase field models for phase decomposition and gas bubble formation, and thermal conductivity for U3Si2 fuel, and precipitation in FeCrAl cladding. The accomplishments are very useful by providing atomistic and mesoscale tools, improving the current understanding, and delivering engineering scale models for these two ATF concepts.

  9. ACCIDENT PREDICTION MODELS FOR UNSIGNALISED URBAN JUNCTIONS IN GHANA

    OpenAIRE

    Mohammed SALIFU, MSc., PhD, MIHT, MGhIE

    2004-01-01

    The main objective of this study was to provide an improved method for safety appraisal in Ghana through the development and application of suitable accident prediction models for unsignalised urban junctions. A case study was designed comprising 91 junctions selected from the two most cosmopolitan cities in Ghana. A wide range of traffic and road data together with the corresponding accident data for each junction for the three-year period 1996-1998 was utilized in the model development p...

  10. Modeling the economic consequences of LWR accidents

    International Nuclear Information System (INIS)

    Burke, R.P.; Aldrich, D.C.; Rasmussen, N.C.

    1984-01-01

    Models to be used for analyses of economic risks from events which may occur during LWR plant operation are developed in this study. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The models can be used by both the nuclear power industry and regulatory agencies in cost-benefit analyses for decisionmaking purposes. The newly developed economic consequence models are applied in an example to estimate the economic risks from operation of the Surry Unit 2 plant. The analyses indicate that economic risks from US LWR operation, in contrast to public health risks, are dominated by relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for the Surry site. The implications of these conclusions for nuclear power plant operation and regulation are discussed

  11. The development of severe accident analysis technology

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-07-01

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

  12. Developments concerning reactivity accidents in PWRs

    International Nuclear Information System (INIS)

    Gouffon, A.

    1987-11-01

    After placing the development work on reactivity accidents in the various actions decided upon further to the Chernobyl accident, this note describes the first results obtained and the further developments. As a general rule, the Chernobyl accident has not provided, from a strictly technical viewpoint, any fundamentally new material which had previouly been unknown. Analysis have made it possible to more clearly establish the safety importance of certain operating rules, in particular concerning handling whithin coolant system pumps. They have not show the need to modify the design of the french PWR.s. This development work must be continued to gain a fuller understanding of the behaviour of fuel, specially after irradiation and power cycling

  13. Longitudinal relationship between economic development and occupational accidents in China.

    Science.gov (United States)

    Song, Li; He, Xueqiu; Li, Chengwu

    2011-01-01

    The relativity between economic development and occupational accidents is a debated topic. Compared with the development courses of both economic development and occupational accidents in China during 1953-2008, this paper used statistic methods such as Granger causality test, cointegration test and impulse response function based on the vector autoregression model to investigate the relativity between economic development and occupational accidents in China from 1953 to 2008. Owing to fluctuation and growth scale characteristics of economic development, two dimensions including economic cycle and economic scale were divided. Results showed that there was no relationship between occupational accidents and economic scale during 1953-1978. Fatality rate per 10(5) workers was a conductive variable to gross domestic product per capita during 1979-2008. And economic cycle was an indicator to occupational accidents during 1979-2008. Variation of economic speed had important influence on occupational accidents in short term. Thus it is necessary to adjust Chinese occupational safety policy according to tempo variation of economic growth. Crown Copyright © 2010. Published by Elsevier Ltd. All rights reserved.

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

    International Nuclear Information System (INIS)

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

    2007-01-01

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

  15. Phased Development of Accident Tolerant Fue

    Energy Technology Data Exchange (ETDEWEB)

    Bragg-Sitton, Shannon M.; Carmack, W. Jon

    2016-09-01

    The United States Department of Energy (U.S. DOE) Advanced Fuels Campaign (AFC) has adopted a three-phase approach for the development and eventual commercialization of enhanced, accident tolerant fuel (ATF) for light water reactors (LWRs). Extending from 2012 to 2016, AFC is currently coming to the end of Phase 1 research that has entailed Feasibility Assessment and Prioritization for a large number of proposed fuel systems (fuel and cladding) that could provide improved performance under accident conditions. Phase 1 activities will culminate with a prioritization of concepts for both near-term and long-term development based on the available experimental data and modeling predictions. This process will provide guidance to DOE on what concepts should be prioritized for investment in Phase 2 Development/Qualification activities based on technical performance improvements and probability of meeting the aggressive schedule to insert a lead fuel rod (LFR) in a commercial power reactor by 2022. While Phase 1 activities include small-scale fabrication work, materials characterization, and limited irradiation of samples, Phase 2 will require development teams to expand to industrial fabrication methods, conduct irradiation tests under more prototypic reactor conditions (i.e. in contact with reactor primary coolant at LWR conditions and in-pile transient testing), conduct additional characterization and post-irradiation examination, and develop a fuel performance code for the candidate ATF. Phase 2 will culminate in the insertion of an LFR (or lead fuel assembly) in a commercial power reactor. The Phase 3 Commercialization work will extend past 2022. Following post-irradiation examination of LFRs, partial-core reloads will be demonstrated. The commercialization phase will further entail the establishment of commercial fabrication capabilities and the transition of LWR cores to the new fuel. The three development phases described roughly correspond to the technology

  16. Economic development, mobility and traffic accidents in Algeria.

    Science.gov (United States)

    Bougueroua, M; Carnis, L

    2016-07-01

    The aim of this contribution is to estimate the impact of road economic conditions and mobility on traffic accidents for the case of Algeria. Using the cointegration approach and vector error correction model (VECM), we will examine simultaneously short term and long-term impacts between the number of traffic accidents, fuel consumption and gross domestic product (GDP) per capital, over the period 1970-2013. The main results of the estimation show that the number of traffic accidents in Algeria is positively influenced by the GDP per capita in the short and long term. It implies that a higher economic development worsens the road safety situation. However, the new traffic rules adopted in 2009 have an impact on the forecast trend of traffic accidents, meaning efficient public policy could improve the situation. This result calls for a strong political commitment with effective countermeasures for avoiding the further deterioration of road safety record in Algeria. Copyright © 2016 Elsevier Ltd. All rights reserved.

  17. Analysis of Accident Scenarios for the Development of Probabilistic Safety Assessment Model for the Metallic Fuel Sodium-Cooled Fast Reactor

    International Nuclear Information System (INIS)

    Kim, Tae Woon; Park, S. Y.; Yang, J. E.; Kwon, Y. M.; Jeong, H. Y.; Suk, S. D.; Lee, Y. B.

    2009-03-01

    The safety analysis reports which were reported during the development of sodium cooled fast reactors in the foreign countries are reviewed for the establishment of Probabilistic Safety Analysis models for the domestic SFR which are under development. There are lots of differences in the safety characteristics between the mixed oxide (MOX) fuel SFR and metallic fuel SFR. Metallic fuel SFR is under development in Korea while MOX fuel SFR is under development in France, Japan, India and China. Therefore the status on the development of fast reactors in the foreign countries are reviewed at first and then the safety characteristics between the MOX fuel SFR and the metallic fuel SFR are reviewed. The core damage can be defined as coolant voiding, fuel melting, cladding damage. The melting points of metallic fuel and the MOX fuel is about 1000 .deg. C and 2300 .deg. C, respectively. The high energy stored in the MOX fuel have higher potential to voiding of coolant compared to the possibility in the metallic fuel. The metallic fuel has also inherent reactivity feedback characteristic that the metallic fuel SFR can be shutdown safely in the events of transient overpower, loss of flow, and loss of heat sink without scram. The metallic fuel has, however, lower melting point due to the eutectic formation between the uranium in metallic fuel and the ferrite in metallic cladding. It is needed to identify the core damage accident scenarios to develop Level-1 PSA model. SSC-K computer code is used to identify the conditions in which the core damage can occur in the KALIMER-600 SFR. The accident cases which are analyzed are the triple failure accidents such as unprotected transient over power events, loss of flow events, and loss of heat sink events with impaired safety systems or functions. Through the analysis of the triple failure accidents for the KALIMER-600 SFR, it is found that the PSA model developed for the PRISM reactor design can be applied to KALIMER-600. However

  18. Development of a kinetic model, including rate constant estimations, on iodine and caesium behaviour in the primary circuit of LWR's under accident conditions

    International Nuclear Information System (INIS)

    Alonso, A.; Buron, J.M.; Fernandez, S.

    1991-07-01

    In this report, a kinetic model has been developed with the aim to try to reproduce the chemical phenomena that take place in a flowing system containing steam, hydrogen and iodine and caesium vapours. The work is divided into two different parts. The first part consists in the estimation, through the Activited Complex Theory, of the reaction rate constants, for the chosen reactions, and the development of the kinetic model based on the concept of ideal tubular chemical reactor. The second part deals with the application of such model to several cases, which were taken from the Phase B 'Scoping Calculations' of the Phebus-FP Project (sequence AB) and the SFD-ST and SFD1.1 experiments. The main conclusion obtained from this work is that the assumption of instantaneous equilibrium could be inacurrate in order to estimate the iodine and caesium species distribution under severe accidents conditions

  19. An application of probabilistic safety assessment methods to model aircraft systems and accidents

    Energy Technology Data Exchange (ETDEWEB)

    Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

    1998-08-01

    A case study modeling the thrust reverser system (TRS) in the context of the fatal accident of a Boeing 767 is presented to illustrate the application of Probabilistic Safety Assessment methods. A simplified risk model consisting of an event tree with supporting fault trees was developed to represent the progression of the accident, taking into account the interaction between the TRS and the operating crew during the accident, and the findings of the accident investigation. A feasible sequence of events leading to the fatal accident was identified. Several insights about the TRS and the accident were obtained by applying PSA methods. Changes proposed for the TRS also are discussed.

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

    International Nuclear Information System (INIS)

    Basic, I.; Kocnar, R.

    1996-01-01

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

  1. Development of an anthropomorphic model and a Monte Carlo calculation code devoted to the physical reconstruction of a radiological accident

    International Nuclear Information System (INIS)

    Roux, A.

    2001-01-01

    The diversity of radiological accidents makes difficult the medical prognosis and the therapy choice from only clinical observations. To complete this information, it is important to know the global dose received by the organism and the dose distributions in depth in tissues. The dose estimation can be made by a physical reconstruction of the accident with the help of tools based on experimental techniques or on calculation. The software of the geometry construction (M.G.E.D.), associated to the Monte-Carlo code of photons and neutrons transport (M.O.R.S.E.) replies these constraints. An important result of this work is to determine the principal parameters to know in function of the accident type, as well as the precision level required for these parameters. (N.C.)

  2. Development of the criticality accident analysis code, AGNES

    International Nuclear Information System (INIS)

    Nakajima, Ken

    1989-01-01

    In the design works for the facilities which handle nuclear fuel, the evaluation of criticality accidents cannot be avoided even if their possibility is as small as negligible. In particular in the system using solution fuel like uranyl nitrate, solution has the property easily becoming dangerous form, and all the past criticality accidents occurred in the case of solution, therefore, the evaluation of criticality accidents becomes the most important item of safety analysis. When a criticality accident occurred in a solution fuel system, due to the generation and movement of radiolysis gas voids, the oscillation of power output and pressure pulses are observed. In order to evaluate the effect of criticality accidents, these output oscillation and pressure pulses must be calculated accurately. For this purpose, the development of the dynamic characteristic code AGNES (Accidentally Generated Nuclear Excursion Simulation code) was carried out. The AGNES is the reactor dynamic characteristic code having two independent void models. Modified energy model and pressure model, and as the benchmark calculation of the AGNES code, the results of the experimental analysis on the CRAC experiment are reported. (K.I.)

  3. Development of a totally integrated severe accident training system

    International Nuclear Information System (INIS)

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

    2006-01-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  5. Development of Severe Accident Containment Analysis Package

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-10-15

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

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

    International Nuclear Information System (INIS)

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

    1986-01-01

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

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

    International Nuclear Information System (INIS)

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

    2005-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-03-15

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

  9. An approach to accidents modeling based on compounds road environments.

    Science.gov (United States)

    Fernandes, Ana; Neves, Jose

    2013-04-01

    The most common approach to study the influence of certain road features on accidents has been the consideration of uniform road segments characterized by a unique feature. However, when an accident is related to the road infrastructure, its cause is usually not a single characteristic but rather a complex combination of several characteristics. The main objective of this paper is to describe a methodology developed in order to consider the road as a complete environment by using compound road environments, overcoming the limitations inherented in considering only uniform road segments. The methodology consists of: dividing a sample of roads into segments; grouping them into quite homogeneous road environments using cluster analysis; and identifying the influence of skid resistance and texture depth on road accidents in each environment by using generalized linear models. The application of this methodology is demonstrated for eight roads. Based on real data from accidents and road characteristics, three compound road environments were established where the pavement surface properties significantly influence the occurrence of accidents. Results have showed clearly that road environments where braking maneuvers are more common or those with small radii of curvature and high speeds require higher skid resistance and texture depth as an important contribution to the accident prevention. Copyright © 2013 Elsevier Ltd. All rights reserved.

  10. Modeling of corium dispersion in DCH accidents

    International Nuclear Information System (INIS)

    Wu, Q.

    1996-01-01

    A model that governs the dispersion process in the direct containment heating (DCH) reactor accident scenario is developed by a stepwise approach. In this model, the whole transient is subdivided into four phases with an isothermal assumption. These are the liquid and gas discharge, the liquid film flow in the cavity before gas blowdown, the liquid and gas flow in the cavity with droplet entrainment, and the liquid transport and re-entrainment in the subcompartment. In each step, the dominant driving mechanisms are identified to construct the governing equations. By combining all the steps together, the corium dispersion information is obtained in detail. The key parameters are predicted quantitatively. These include the fraction of liquid that flows out of the cavity before gas blowdown, the dispersion fraction and the mean droplet diameter in the cavity, the cavity pressure rise due to the liquid friction force, and the dispersion fractions in the containment via different paths. Compared with the data of the 1:10 scale experiments carried out at Purdue University, fairly good agreement is obtained. A stand-alone prediction of the corium dispersion under prototypic Zion reactor conditions is carried out by assuming an isothermal process without chemical reactions. (orig.)

  11. PSA modeling of long-term accident sequences

    International Nuclear Information System (INIS)

    Georgescu, Gabriel; Corenwinder, Francois; Lanore, Jeanne-Marie

    2014-01-01

    In the context of the extension of PSA scope to include external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the accident sequences induced by initiators which affect a whole site containing several nuclear units (reactors, fuel pools,...). These methodological improvements represent an essential prerequisite for the development of external hazards PSA. However, it has to be noted that in French PSA, even before Fukushima, long term accident sequences were taken into account: many insight were therefore used, as complementary information, to enhance the safety level of the plants. IRSN proposed an external events PSA development program. One of the first steps of the program is the development of methods to model in the PSA the long term accident sequences, based on the experience gained. At short term IRSN intends to enhance the modeling of the 'long term' accident sequences induced by the loss of the heat sink or/and the loss of external power supply. The experience gained by IRSN and EDF from the development of several probabilistic studies treating long term accident sequences shows that the simple extension of the mission time of the mitigation systems from 24 hours to longer times is not sufficient to realistically quantify the risk and to obtain a correct ranking of the risk contributions and that treatment of recoveries is also necessary. IRSN intends to develop a generic study which can be used as a general methodology for the assessment of the long term accident sequences, mainly generated by external hazards and their combinations. This first attempt to develop this generic study allowed identifying some aspects, which may be hazard (or combinations of hazards) or related to initial boundary conditions, which should be taken into account for further developments. (authors)

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

    International Nuclear Information System (INIS)

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

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  14. Development of Integrated Evaluation System for Severe Accident Management

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-06-15

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

  15. Development of Integrated Evaluation System for Severe Accident Management

    International Nuclear Information System (INIS)

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

    2007-06-01

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

  16. Development of a model to predict flow oscillations in low-flow sodium boiling. [Loss-of-Piping Integrity accidents

    Energy Technology Data Exchange (ETDEWEB)

    Levin, A.E.; Griffith, P.

    1980-04-01

    Tests performed in a small scale water loop showed that voiding oscillations, similar to those observed in sodium, were present in water, as well. An analytical model, appropriate for either sodium or water, was developed and used to describe the water flow behavior. The experimental results indicate that water can be successfully employed as a sodium simulant, and further, that the condensation heat transfer coefficient varies significantly during the growth and collapse of vapor slugs during oscillations. It is this variation, combined with the temperature profile of the unheated zone above the heat source, which determines the oscillatory behavior of the system. The analytical program has produced a model which qualitatively does a good job in predicting the flow behavior in the wake experiment. The amplitude discrepancies are attributable to experimental uncertainties and model inadequacies. Several parameters (heat transfer coefficient, unheated zone temperature profile, mixing between hot and cold fluids during oscillations) are set by the user. Criteria for the comparison of water and sodium experiments have been developed.

  17. Accident consequence assessment and siting criteria development

    International Nuclear Information System (INIS)

    Kollas, J.G.

    1988-01-01

    The methodology developed is based on assessing the average over a large spectrum of meteorological conditions whole body collective dose resulting from a severe reference accident. The assessment of this dose is performed by code CRAC.GAEC, the Greek A.E.C. version of code CRAC2. The collective dose, which is chosen as a measure of the social radiation risk, is compared to the dose corresponding to a level of social risk encountered historically in energy production as a whole. The outcome of the comparison can be the determination of one or more sectors of acceptable sites for a set of specific conditions considered, such as the reactor characteristics. The present approach was aimed to deal with the problems stemming from the demographic idiomorphy of Greece, where one third of the country's population is concentrated in Athens, with the rest of the country exhibiting small population densities. One of the applications of the methodology developed concerned the identification of acceptable sites near Athens. For these sites the risk from the reference severe accident of a standard reactor to the over three millions inhabitants of Athens is less tan the risk corresponding to the same population that is due to energy production

  18. Response of a DSNP pressurizer model under accident conditions

    International Nuclear Information System (INIS)

    Saphier, D.; Kallfelz, J.; Belblidia, L.

    1986-01-01

    Recently a new pressurizer model was developed for the DSNP simulation language. The model was connected to a simulation of the Trojan pressurized water reactor (PWR) and tested by simulating a loss-of-off-site power (LOSP) anticipated transient without scram. The results compare well to a similar study performed using the RELAP code. The pressurizer model and its response to the LOSP accident are presented

  19. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Zhegang Ma

    2013-09-01

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  20. A model national emergency plan for radiological accidents

    International Nuclear Information System (INIS)

    2000-07-01

    The IAEA has supported several projects for the development of a national response plan for radiological emergencies. As a result, the IAEA has developed a model National Emergency Response Plan for Radiological Accidents (RAD PLAN), particularly for countries that have no nuclear power plants. This plan can be adapted for use by countries interested in developing their own national radiological emergency response plan, and the IAEA will supply the latest version of the RAD PLAN on computer diskette upon request

  1. A model national emergency response plan for radiological accidents

    International Nuclear Information System (INIS)

    1993-09-01

    The IAEA has supported several projects for the development of a national response plan for radiological emergencies. As a results, the IAEA has developed a model National Emergency Response Plan for Radiological Accidents (RAD PLAN), particularly for countries that have no nuclear power plants. This plan can be adapted for use by countries interested in developing their own national radiological emergency response plan, and the IAEA will supply the latest version of the RAD PLAN on computer diskette upon request. 2 tabs

  2. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented...

  3. Applying Functional Modeling for Accident Management of Nucler Power Plant

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role...... for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented....

  4. WASA-BOSS. Development and application of Severe Accident Codes. Evaluation and optimization of accident management measures. Subproject E. Improvement of the lower head model in MELCOR and calculations in connection with the FUKUSHIMA accident. Final report; WASA-BOSS. Weiterentwicklung und Anwendung von Severe Accident Codes. Bewertung und Optimierung von Stoerfallmassnahmen. Teilprojekt E. Verbesserung des Lower Head-Modelles fuer MELCOR und MELCOR-Rechnungen zu Fukushima. Schlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Kretzschmar, Frank; Dietrich, Philipp; Gabriel, Stephan; Miassoedov, Alexei

    2016-12-15

    The knowledge of the key phenomena, which govern the chronological sequence of a core melt accident, is a crucial precondition for the development of SAMGs (Severe Management Guides) to avoid and mitigate the radiological consequences for the population and the environment. In the frame of a dissertation a new model has been coupled with MELCOR, which describes the thermal interaction of a core melt with the RPV (reactor pressure vessel) wall in the lower plenum. This model allows a better description of this phenomenon. The method to couple extern programs with MELCOR had been already developed and used in a former dissertation at KIT-IKET. The model has been validated recalculating according experiments in the LIVE facility. Afterwards a defined accident scenario has been calculated for a German generic KONVOI power plant. 12 months after the start of the project a MELCOR input has been developed using data delivered by the Ruhr university of Bochum (subproject ''Simulation des Unfalls in Fukushima-Daichi zur Bewertung des Stoerfall-Analysecodes ATHLET-CD''). The results of this simulation have made a contribution to review the current understanding of the FUKUSHIMA sequence. HZDR and KIT-IKET have agreed in the course of the project, that KIT-IKET will develop a MELCOR input of a german generic KONVOI power plant following an ATHLET-CDinput of HZDR. Using this MELCOR input, a comparative analysis has been performed.

  5. Development of INCTAC code for analyzing criticality accident phenomena

    International Nuclear Information System (INIS)

    Mitake, Susumu; Hayashi, Yamato; Sakurai, Shungo

    2003-01-01

    Aiming at understanding nuclear transients and thermal- and hydraulic-phenomena of the criticality accident, a code named INCTAC has been newly developed at the Institute of Nuclear Safety. The code is applicable to the analysis of criticality accident transients of aqueous homogenous fuel solution system. Neutronic transient model is composed of equations for the kinetics and for the spatial distributions, which are deduced from the time dependent multi-group transport equations with the quasi steady state assumption. Thermal-hydraulic transient model is composed of a complete set of the mass, momentum and energy equations together with the two-phase flow assumptions. Validation tests of INCTAC were made using the data obtained at TRACY, a transient experiment criticality facility of JAERI. The calculated results with INCTAC showed a very good agreement with the experiment data, except a slight discrepancy of the time when the peak of reactor power was attained. But, the discrepancy was resolved with the use of an adequate model for movement and transfer of the void in the fuel solution mostly generated by radiolysis. With a simulation model for the transport of radioactive materials through ventilation systems to the environment, INCTAC will be used as an overall safety evaluation code of the criticality accident. (author)

  6. The effect of system modeling on the Fukushima accident evolution

    Energy Technology Data Exchange (ETDEWEB)

    Herranz, L.E.; Fontanet, J.; López, C.; Fernández, E.

    2015-07-01

    The Fukushima accident is becoming both a unique opportunity and a huge challenge for severe accident analysis. The OECD-BSAF project has articulated a good part of the modeling efforts conducted so far. Inside this project, CIEMAT has conducted forensic analyses of the Fukushima accident in units 1 through 3 with MELCOR 2.1 and it has postulated a set of accident scenarios consistent with data. Beyond specific results, sensitivity analyses on safety systems performance and prevailing boundary conditions have highlighted the need of conducting uncertainty analyses when modeling NPPs severe accident scenarios. (Author)

  7. MELCOR modeling of Fukushima unit 2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Sevon, Tuomo [VTT Technical Research Centre of Finland, Espoo (Finland)

    2014-12-15

    A MELCOR model of the Fukushima Daiichi unit 2 accident was created in order to get a better understanding of the event and to improve severe accident modeling methods. The measured pressure and water level could be reproduced relatively well with the calculation. This required adjusting the RCIC system flow rates and containment leak area so that a good match to the measurements is achieved. Modeling of gradual flooding of the torus room with water that originated from the tsunami was necessary for a satisfactory reproduction of the measured containment pressure. The reactor lower head did not fail in this calculation, and all the fuel remained in the RPV. 13 % of the fuel was relocated from the core area, and all the fuel rods lost their integrity, releasing at least some volatile radionuclides. According to the calculation, about 90 % of noble gas inventory and about 0.08 % of cesium inventory was released to the environment. The release started 78 h after the earthquake, and a second release peak came at 90 h. Uncertainties in the calculation are very large because there is scarce public data available about the Fukushima power plant and because it is not yet possible to inspect the status of the reactor and the containment. Uncertainty in the calculated cesium release is larger than factor of ten.

  8. Developing and assessing accident management plans for nuclear power plants

    International Nuclear Information System (INIS)

    Hanson, D.J.; Johnson, S.P.; Blackman, H.S.; Stewart, M.A.

    1992-07-01

    This document is the second of a two-volume NUREG/CR that discusses development of accident management plans for nuclear power plants. The first volume (a) describes a four-phase approach for developing criteria that could be used for assessing the adequacy of accident management plans, (b) identifies the general attributes of accident management plans (Phase 1), (c) presents a prototype process for developing and implementing severe accident management plans (Phase 2), and (d) presents criteria that can be used to assess the adequacy of accident management plans. This volume (a) describes results from an evaluation of the capabilities of the prototype process to produce an accident management plan (Phase 3) and (b), based on these results and preliminary criteria included in NUREG/CR-5543, presents modifications to the criteria where appropriate

  9. Development of radiation dose assessment system for radiation accident (RADARAC)

    International Nuclear Information System (INIS)

    Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

    2009-07-01

    The possibility of radiation accident is very rare, but cannot be regarded as zero. Medical treatments are quite essential for a heavily exposed person in an occurrence of a radiation accident. Radiation dose distribution in a human body is useful information to carry out effectively the medical treatments. A radiation transport calculation utilizing the Monte Carlo method has an advantageous in the analysis of radiation dose inside of the body, which cannot be measured. An input file, which describes models for the accident condition and quantities of interest, should be prepared to execute the radiation transport calculation. Since the accident situation, however, cannot be prospected, many complicated procedures are needed to make effectively the input file soon after the occurrence of the accident. In addition, the calculated doses are to be given in output files, which usually include much information concerning the radiation transport calculation. Thus, Radiation Dose Assessment system for Radiation Accident (RADARAC) was developed to derive effectively radiation dose by using the MCNPX or MCNP code. RADARAC mainly consists of two parts. One part is RADARAC - INPUT, which involves three programs. A user can interactively set up necessary resources to make input files for the codes, with graphical user interfaces in a personnel computer. The input file includes information concerning the geometric structure of the radiation source and the exposed person, emission of radiations during the accident, physical quantities of interest and so on. The other part is RADARAC - DOSE, which has one program. The results of radiation doses can be effectively indicated with numerical tables, graphs and color figures visibly depicting dose distribution by using this program. These results are obtained from the outputs of the radiation transport calculations. It is confirmed that the system can effectively make input files with a few thousand lines and indicate more than 20

  10. Developement of integrated evaluation system for severe accident management

    Energy Technology Data Exchange (ETDEWEB)

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

    2005-04-01

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

  11. Investigation of Zircaloy-2 oxidation model for SFP accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    Nemoto, Yoshiyuki, E-mail: nemoto.yoshiyuki@jaea.go.jp [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Kaji, Yoshiyuki; Ogawa, Chihiro; Kondo, Keietsu [Japan Atomic Energy Agency, 2-4 Shirakata, Ohaza, Tokai-mura, Naka-gun, Ibaraki, 319-1195 (Japan); Nakashima, Kazuo; Kanazawa, Toru; Tojo, Masayuki [Global Nuclear Fuel – Japan Co., Ltd., 2-3-1, Uchikawa, Yokosuka-shi, Kanagawa, 239-0836 (Japan)

    2017-05-15

    The authors previously conducted thermogravimetric analyses on Zircaloy-2 in air. By using the thermogravimetric data, an oxidation model was constructed in this study so that it can be applied for the modeling of cladding degradation in spent fuel pool (SFP) severe accident condition. For its validation, oxidation tests of long cladding tube were conducted, and computational fluid dynamics analyses using the constructed oxidation model were proceeded to simulate the experiments. In the oxidation tests, high temperature thermal gradient along the cladding axis was applied and air flow rates in testing chamber were controlled to simulate hypothetical SFP accidents. The analytical outputs successfully reproduced the growth of oxide film and porous oxide layer on the claddings in oxidation tests, and validity of the oxidation model was proved. Influence of air flow rate for the oxidation behavior was thought negligible in the conditions investigated in this study. - Highlights: •An oxidation model of Zircaloy-2 in air environment was developed. •The oxidation model was validated by the comparison with oxidation tests using long cladding tubes in hypothetical spent fuel pool accident condition. •The oxidation model successfully reproduced the typical oxidation behavior in air.

  12. Design and Development of a Severe Accident Training System

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

    Ahn, Kwang Il; Kim, Dong Ha

    2003-01-01

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

  15. Use of flow models to analyse loss of coolant accidents

    International Nuclear Information System (INIS)

    Pinet, Bernard

    1978-01-01

    This article summarises current work on developing the use of flow models to analyse loss-of-coolant accident in pressurized-water plants. This work is being done jointly, in the context of the LOCA Technical Committee, by the CEA, EDF and FRAMATOME. The construction of the flow model is very closely based on some theoretical studies of the two-fluid model. The laws of transfer at the interface and at the wall are tested experimentally. The representativity of the model then has to be checked in experiments involving several elementary physical phenomena [fr

  16. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    Togawa, O.; Homma, T.

    1992-01-01

    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)

  17. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2016-10-01

    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.

  18. Developments pertaining to the Three Mile Island accident

    International Nuclear Information System (INIS)

    Cottrell, W.B.

    1979-01-01

    Some of the developments since August 1979 pertaining to the accident are of general interest and are summarized. No attempt is made to present a comprehensive review of the accident nor to evaluate the material that has become available. The article merely calls attention to the available information

  19. Development of a methodology for accident causation research

    Science.gov (United States)

    1983-06-01

    The obj ective of this study was to fully develop and apply a me thodology to : study accident causation, uhich was outlined in a previous study . " Causal" factors : are those pre-crash factors, which are statistically related to the accident rate :...

  20. Development of an accident management expert system for containment assessment

    International Nuclear Information System (INIS)

    Nelson, W.R.; Sebo, D.E.; Haney, L.N.

    1987-01-01

    The United States Nuclear Regulatory Commission (USNRSC) is sponsoring a program at the Idaho National Engineering Laboratory (INEL) to develop an accident management expert system. The intended users of the system are the personnel of the NRC Operations Center in Washington, D.C. The expert system will be used to help NRC personnel monitor and evaluate the status and management of the containment during a severe reactor accident. The knowledge base will include severe accident knowledge regarding the maintenance of the critical safety functions, especially containment integrity, during an accident. This paper summarizes the concepts that have been developed for the accident management expert system, and the plans that have been developed for its implementation

  1. Dust mobilization and transport modeling for loss of vacuum accidents

    International Nuclear Information System (INIS)

    Humrickhouse, P.W.; Sharpe, J.P.

    2007-01-01

    We develop a general continuum fluid dynamic model for dust transport in loss of vacuum accidents in fusion energy systems. The relationship between this general approach and established particle transport methods is clarified, in particular the relationship between the seemingly disparate treatments of aerosol dynamics and Lagrangian particle tracking. Constitutive equations for granular flow are found to be inadequate for prediction of mobilization, as these models essentially impose a condition of flow from the outset. Experiments confirm that at low shear, settled dust piles behave more like a continuum solid, and suitable solid models will be required to predict the onset of dust mobilization

  2. Dust mobilization and transport modeling for loss of vacuum accidents

    International Nuclear Information System (INIS)

    Humrickhouse, P.W.; Sharpe, J.P.

    2008-01-01

    We develop a general continuum fluid dynamic model for dust transport in loss of vacuum accidents in fusion energy systems. The relationship between this general approach and established particle transport methods is clarified, in particular the relationship between the seemingly disparate treatments of aerosol dynamics and Lagrangian particle tracking. Constitutive equations for granular flow are found to be inadequate for prediction of mobilization, as these models essentially impose a condition of flow from the outset. Experiments confirm that at low shear, settled dust piles behave more like a continuum solid, and suitable solid models will be required to predict the onset of dust mobilization

  3. A cascading failure model for analyzing railway accident causation

    Science.gov (United States)

    Liu, Jin-Tao; Li, Ke-Ping

    2018-01-01

    In this paper, a new cascading failure model is proposed for quantitatively analyzing the railway accident causation. In the model, the loads of nodes are redistributed according to the strength of the causal relationships between the nodes. By analyzing the actual situation of the existing prevention measures, a critical threshold of the load parameter in the model is obtained. To verify the effectiveness of the proposed cascading model, simulation experiments of a train collision accident are performed. The results show that the cascading failure model can describe the cascading process of the railway accident more accurately than the previous models, and can quantitatively analyze the sensitivities and the influence of the causes. In conclusion, this model can assist us to reveal the latent rules of accident causation to reduce the occurrence of railway accidents.

  4. Modelling road accidents: An approach using structural time series

    Science.gov (United States)

    Junus, Noor Wahida Md; Ismail, Mohd Tahir

    2014-09-01

    In this paper, the trend of road accidents in Malaysia for the years 2001 until 2012 was modelled using a structural time series approach. The structural time series model was identified using a stepwise method, and the residuals for each model were tested. The best-fitted model was chosen based on the smallest Akaike Information Criterion (AIC) and prediction error variance. In order to check the quality of the model, a data validation procedure was performed by predicting the monthly number of road accidents for the year 2012. Results indicate that the best specification of the structural time series model to represent road accidents is the local level with a seasonal model.

  5. Developing a comprehensive and accountable database after a radiological accident

    International Nuclear Information System (INIS)

    Berry, H.A.; Burson, Z.G.

    1986-09-01

    After a radiological accident occurs, it is highly desirable to promptly begin developing a comprehensive and accountable environmental database both for immediate health and safety needs and for long-term documentation. The need to assess and evaluate the impact of the accident as quickly as possible is always very urgent, the technical integrity of the data must also be assured and maintained. Care must therefore be taken to log, collate, and organize the environmental data into a complete and accountable database. The key components of the database development are summarized as well as the experience gained in organizing and handling environmental data acquired during: (1) TMI (1979); (2) the St. Lucie Reactor Accident Exercise (through the Federal Radiological Measurement and Assessment Center (FRMAC), March 1984); (3) the Sequoyah Fuels Inc., uranium hexafluoride accident near Gore, Oklahoma (January 1986); and (4) Chernobyl reactor accident in Russia (April 1986)

  6. Modeling atmospheric dispersion for reactor accident consequence evaluation

    International Nuclear Information System (INIS)

    Alpert, D.J.; Gudiksen, P.H.; Woodard, K.

    1982-01-01

    Atmospheric dispersion models are a central part of computer codes for the evaluation of potential reactor accident consequences. A variety of ways of treating to varying degrees the many physical processes that can have an impact on the predicted consequences exists. The currently available models are reviewed and their capabilities and limitations, as applied to reactor accident consequence analyses, are discussed

  7. Modeling of criticality accidents and their environmental consequences

    International Nuclear Information System (INIS)

    Thomas, W.; Gmal, B.

    1987-01-01

    In the Federal Republic of Germany, potential radiological consequences of accidental nuclear criticality have to be evaluated in the licensing procedure for fuel cycle facilities. A prerequisite to this evaluation is to establish conceivable accident scenarios. First, possibilities for a criticality exceeding the generally applied double contingency principle of safety are identified by screening the equipment and operation of the facility. Identification of undetected accumulations of fissile material or incorrect transfer of fissile solution to unfavorable geometry normally are most important. Second, relevant and credible scenarios causing the most severe consequences are derived from these possibilities. For the identified relevant scenarios, time-dependent fission rates and reasonable numbers for peak power and total fissions must be determined. Experience from real accidents and experiments (KEWB, SPERT, CRAC, SILENE) has been evaluated using empirical formulas. To model the time-dependent behavior of criticality excursions in fissile solutions, a computer program FELIX has been developed

  8. Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

    International Nuclear Information System (INIS)

    Kim, Bo Gyung; Kang Hyun Gook; Yoon, Ho Joon

    2016-01-01

    In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results

  9. Sequence Tree Modeling for Combined Accident and Feed-and-Bleed Operation

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Bo Gyung; Kang Hyun Gook [KAIST, Daejeon (Korea, Republic of); Yoon, Ho Joon [Khalifa University of Science, Abu Dhabi (United Arab Emirates)

    2016-05-15

    In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model can translate into a dynamic event tree model based on the sampling analysis results.

  10. Public transportation development and traffic accident prevention in Indonesia

    Directory of Open Access Journals (Sweden)

    Sutanto Soehodho

    2017-03-01

    Full Text Available Traffic accidents have long been known as an iceberg for comprehending the discrepancies of traffic management and entire transportation systems. Figures detailing traffic accidents in Indonesia, as is the case in many other countries, show significantly high numbers and severity levels; these types of totals are also evident in Jakarta, the highest-populated city in the country. While the common consensus recognizes that traffic accidents are the results of three different factor types, namely, human factors, vehicle factors, and external factors (including road conditions, human factors have the strongest influence—and figures on a worldwide scale corroborate that assertion. We, however, try to pinpoint the issues of non-human factors in light of increasing traffic accidents in Indonesia, where motorbike accidents account for the majority of incidents. We then consider three important pillars of action: the development of public transportation, improvement of the road ratio, and traffic management measures.

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

    International Nuclear Information System (INIS)

    Kaliatka, A.; Valinčius, M.

    2012-01-01

    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.

  12. Development of PSA module for computerized accident management support (CAMS)

    International Nuclear Information System (INIS)

    Iguchi, Yukihiro

    1996-10-01

    CAMS (Computerised Accident Management Support) is a system that will provide assistance in case of the accidents in a nuclear power plant. The PSA module was developed in order to give useful information in this situation applying the PSA method, which is a comprehensive source of safety knowledge. This module contains plant-specific PSA data, comprising event trees, failure probabilities etc. It has several event trees categorised according to the initiating events. Each event tree has an initiating event frequency and branching probabilities. The various support systems for branches are considered and their dependencies are calculated logically. This module can be activated by data from the state identification (SI) module of CAMS. If an initiating event occurs, the event tree is re-calculated and the PSA module shows which systems of the plant should be activated to bring the plant to a safe state. If the plant responds to the event in the normal way, the plant will be shut down and come to a safe state. However, if some functions do not work, the PSA module generates another path and gives information about the critical systems. If the state of the plant is changed, either by the operators or automatically by the control system, the PSA module follows the new path. Because the estimation of the core damage frequency should be very quick in the accident situation, a simplified model of the event tree and fault trees was adopted. It enabled the PSA module to calculates the CDF within 5 seconds on a standard type work station. The development of the module has been successful. However, further development of the functionality of the module is suggested like real connection to a plant and to the strategy generator module of CAMS, applications for operational support, low power operation optimisation, etc. (author)

  13. Development of Traffic Accidents Control System

    Directory of Open Access Journals (Sweden)

    Andrey Borisovich Nikolaev

    2015-05-01

    Full Text Available Proposed a structure of traffic accidents control system included three main parts: pre-processing, decision support and monitoring. For decision support systems we propose a method that allows to make decisions on the basis of fuzzy situational management. The advantage of the method: it allows to formalize a set of typical traffic situations, using the theory of fuzzy sets and to carry out selection of the desired management action.

  14. Accident simulator development for probabilistic safety analysis

    International Nuclear Information System (INIS)

    Cacciabue, P.C.; Amendola, A.; Mancini, G.

    1985-01-01

    This paper describes the basic features of a new concept of incident simulator, Response System Analyzed (RSA) which is being developed within the CEC JRC Research Program on Reactor Safety. Focusing on somewhat different aims than actual simulators, RSA development extends the field of application of simulators to the area of risk and reliability analysis and in particular to the identification of relevant sequences, to the modeling of human behavior and to the validation of operating procedures. The fundamental components of the project, i.e. the deterministic transient model of the plant, the automatic probabilistic driver and the human possible intervention modeling, are discussed in connection with the problem of their dynamic interaction. The analyses so far performed by separately testing RSA on significant study cases have shown encouraging results and have proven the feasibility of the overall program

  15. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    Energy Technology Data Exchange (ETDEWEB)

    Lind, Morten; Zhang Xinxin [Harbin Engineering University, Harbin (China)

    2014-08-15

    The paper investigate applications of functional modeling for accident management in complex industrial plant with special reference to nuclear power production. Main applications for information sharing among decision makers and decision support are identified. An overview of Multilevel Flow Modeling is given and a detailed presentation of the foundational means-end concepts is presented and the conditions for proper use in modelling accidents are identified. It is shown that Multilevel Flow Modeling can be used for modelling and reasoning about design basis accidents. Its possible role for information sharing and decision support in accidents beyond design basis is also indicated. A modelling example demonstrating the application of Multilevel Flow Modelling and reasoning for a PWR LOCA is presented.

  16. Development of Parameter Network for Accident Management Applications

    Energy Technology Data Exchange (ETDEWEB)

    Pak, Sukyoung; Ahemd, Rizwan; Heo, Gyunyoung [Kyung Hee Univ., Yongin (Korea, Republic of); Kim, Jung Taek; Park, Soo Yong; Ahn, Kwang Il [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2013-10-15

    When a severe accident happens, it is hard to obtain the necessary information to understand of internal status because of the failure or damage of instrumentation and control systems. We learned the lessons from Fukushima accident that internal instrumentation system should be secured and must have ability to react in serious conditions. While there might be a number of methods to reinforce the integrity of instrumentation systems, we focused on the use of redundant behavior of plant parameters without additional hardware installation. Specifically, the objective of this study is to estimate the replaced value which is able to identify internal status by using set of available signals when it is impossible to use instrumentation information in a severe accident, which is the continuation of the paper which was submitted at the last KNS meeting. The concept of the VPN was suggested to improve the quality of parameters particularly to be logged during severe accidents in NPPs using a software based approach, and quantize the importance of each parameter for further maintenance. In the future, we will continue to perform the same analysis to other accident scenarios and extend the spectrum of initial conditions so that we are able to get more sets of VPNs and ANN models to predict the behavior of accident scenarios. The suggested method has the uncertainty underlain in the analysis code for severe accidents. However, In case of failure to the safety critical instrumentation, the information from the VPN would be available to carry out safety management operation.

  17. Development of Accident Scenario for Interim Spent Fuel Storage Facility Based on Fukushima Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Dongjin; Choi, Kwangsoon; Yoon, Hyungjoon; Park, Jungsu [KEPCO-E and C, Yongin (Korea, Republic of)

    2014-05-15

    700 MTU of spent nuclear fuel is discharged from nuclear fleet every year and spent fuel storage is currently 70.9% full. The on-site wet type spent fuel storage pool of each NPP(nuclear power plants) in Korea will shortly exceed its storage limit. Backdrop, the Korean government has rolled out a plan to construct an interim spent fuel storage facility by 2024. However, the type of interim spent fuel storage facility has not been decided yet in detail. The Fukushima accident has resulted in more stringent requirements for nuclear facilities in case of beyond design basis accidents. Therefore, there has been growing demand for developing scenario on interim storage facility to prepare for beyond design basis accidents and conducting dose assessment based on the scenario to verify the safety of each type of storage.

  18. The DOE technology development programme on severe accident management

    International Nuclear Information System (INIS)

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

    1998-01-01

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

  19. Modeling framework for crew decisions during accident sequences

    International Nuclear Information System (INIS)

    Lukic, Y.D.; Worledge, D.H.; Hannaman, G.W.; Spurgin, A.J.

    1986-01-01

    The ability to model the average behavior of operating crews in the course of accident sequences is vital in learning on how to prevent damage to power plants and to maintain safety. This paper summarizes the work carried out in support of a Human Reliability Model framework. This work develops the mathematical framework of the model and identifies the parameters which could be measured in some way, e.g., through simulator experience and/or small scale tests. Selected illustrative examples are presented, of the numerical experiments carried out in order to understand the model sensitivity to parameter variation. These examples ar discussed with the objective of deriving insights of general nature regarding operating of the model which may lead to enhanced understanding of man/machine interactions

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

    International Nuclear Information System (INIS)

    Cizel, F.

    1999-01-01

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

  1. Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

    International Nuclear Information System (INIS)

    Lee, Yongjin; Jae, Moosung

    2014-01-01

    The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

  2. Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

    International Nuclear Information System (INIS)

    Chang, Y.H.; Mosleh, A.; Dang, V.N.

    2003-01-01

    The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

  3. Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.H.; Mosleh, A.; Dang, V.N

    2003-03-01

    The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

  4. Accident-generated radioactive particle source term development for consequence assessment of nuclear fuel cycle facilities

    International Nuclear Information System (INIS)

    Sutter, S.L.; Ballinger, M.Y.; Halverson, M.A.; Mishima, J.

    1983-04-01

    Consequences of nuclear fuel cycle facility accidents can be evaluated using aerosol release factors developed at Pacific Northwest Laboratory. These experimentally determined factors are compiled and consequence assessment methods are discussed. Release factors can be used to estimate the fraction of material initially made airborne by postulated accident scenarios. These release fractions in turn can be used in models to estimate downwind contamination levels as required for safety assessments of nuclear fuel cycle facilities. 20 references, 4 tables

  5. Health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Abrahamson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.

    1993-05-01

    The Nuclear Regulatory Commission (NRC) has sponsored several studies to identify and quantify, through the use of models, the potential health effects of accidental releases of radionuclides from nuclear power plants. The Reactor Safety Study provided the basis for most of the earlier estimates related to these health effects. Subsequent efforts by NRC-supported groups resulted in improved health effects models that were published in the report entitled open-quotes Health Effects Models for Nuclear Power Plant Consequence Analysisclose quotes, NUREG/CR-4214, 1985 and revised further in the 1989 report NUREG/CR-4214, Rev. 1, Part 2. The health effects models presented in the 1989 NUREG/CR-4214 report were developed for exposure to low-linear energy transfer (LET) (beta and gamma) radiation based on the best scientific information available at that time. Since the 1989 report was published, two addenda to that report have been prepared to (1) incorporate other scientific information related to low-LET health effects models and (2) extend the models to consider the possible health consequences of the addition of alpha-emitting radionuclides to the exposure source term. The first addendum report, entitled open-quotes Health Effects Models for Nuclear Power Plant Accident Consequence Analysis, Modifications of Models Resulting from Recent Reports on Health Effects of Ionizing Radiation, Low LET Radiation, Part 2: Scientific Bases for Health Effects Models,close quotes was published in 1991 as NUREG/CR-4214, Rev. 1, Part 2, Addendum 1. This second addendum addresses the possibility that some fraction of the accident source term from an operating nuclear power plant comprises alpha-emitting radionuclides. Consideration of chronic high-LET exposure from alpha radiation as well as acute and chronic exposure to low-LET beta and gamma radiations is a reasonable extension of the health effects model

  6. Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

    International Nuclear Information System (INIS)

    Gamble, K. A.; Hales, J. D.; Yu, J.; Zhang, Y.; Bai, X.; Andersson, D.; Patra, A.; Wen, W.; Tome, C.; Baskes, M.; Martinez, E.; Stanek, C. R.; Miao, Y.; Ye, B.; Hofman, G. L.; Yacout, A. M.; Liu, W.

    2015-01-01

    U 3 Si 2 and iron-chromium-aluminum (Fe-Cr-Al) alloys are two of many proposed accident-tolerant fuel concepts for the fuel and cladding, respectively. The behavior of these materials under normal operating and accident reactor conditions is not well known. As part of the Department of Energy's Accident Tolerant Fuel High Impact Problem program significant work has been conducted to investigate the U 3 Si 2 and FeCrAl behavior under reactor conditions. This report presents the multiscale and multiphysics effort completed in fiscal year 2015. The report is split into four major categories including Density Functional Theory Developments, Molecular Dynamics Developments, Mesoscale Developments, and Engineering Scale Developments. The work shown here is a compilation of a collaborative effort between Idaho National Laboratory, Los Alamos National Laboratory, Argonne National Laboratory and Anatech Corp.

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2013-08-15

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

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

    International Nuclear Information System (INIS)

    2013-01-01

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

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

    International Nuclear Information System (INIS)

    Choi, Young; Park, Soo Yong; Ahn, Kwang-Il; Kim, D.H.

    2006-01-01

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

  10. Identifying traffic accident black spots with Poisson-Tweedie models

    DEFF Research Database (Denmark)

    Debrabant, Birgit; Halekoh, Ulrich; Bonat, Wagner Hugo

    2018-01-01

    This paper aims at the identification of black spots for traffic accidents, i.e. locations with accident counts beyond what is usual for similar locations, using spatially and temporally aggregated hospital records from Funen, Denmark. Specifically, we apply an autoregressive Poisson-Tweedie model...... considered calendar years and calculated by simulations a probability of p=0.03 for these to be chance findings. Altogether, our results recommend these sites for further investigation and suggest that our simple approach could play a role in future area based traffic accident prevention planning....

  11. Phenomenological and mechanistic modeling of melt-structure-water interactions in a light water reactor severe accident

    International Nuclear Information System (INIS)

    Bui, V.A.

    1998-01-01

    The objective of this work is to address the modeling of the thermal hydrodynamic phenomena and interactions occurring during the progression of reactor severe accidents. Integrated phenomenological models are developed to describe the accident scenarios, which consist of many processes, while mechanistic modeling, including direct numerical simulation, is carried out to describe separate effects and selected physical phenomena of particular importance

  12. TIRE MODELS USED IN VEHICLE DYNAMIC APPLICATIONS AND THEIR USING IN VEHICLE ACCIDENT SIMULATIONS

    Directory of Open Access Journals (Sweden)

    Osman ELDOĞAN

    1995-01-01

    Full Text Available Wheel model is very important in vehicle modelling, it is because the contact between vehicle and road is achieved by wheel. Vehicle models can be dynamic models which are used in vehicle design, they can also be models used in accident simulations. Because of the importance of subject, many studies including theoretical, experimental and mixed type have been carried out. In this study, information is given about development of wheel modelling and research studies and also use of these modellings in traffic accident simulations.

  13. Guidelines for system modeling: pre-accident human errors, rev.0

    International Nuclear Information System (INIS)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E.

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors

  14. Guidelines for system modeling: pre-accident human errors, rev.0

    Energy Technology Data Exchange (ETDEWEB)

    Kang, Dae Il; Jung, W. D.; Lee, Y. H.; Hwang, M. J.; Yang, J. E

    2004-01-01

    The evaluation results of Human Reliability Analysis (HRA) of pre-accident human errors in the probabilistic safety assessment (PSA) for the Korea Standard Nuclear Power Plant (KSNP) using the ASME PRA standard show that more than 50% of 10 items to be improved are related to the identification and screening analysis for them. Thus, we developed a guideline for modeling pre-accident human errors for the system analyst to resolve some items to be improved for them. The developed guideline consists of modeling criteria for the pre-accident human errors (identification, qualitative screening, and common restoration errors) and detailed guidelines for pre-accident human errors relating to testing, maintenance, and calibration works of nuclear power plants (NPPs). The system analyst use the developed guideline and he or she applies it to the system which he or she takes care of. The HRA analyst review the application results of the system analyst. We applied the developed guideline to the auxiliary feed water system of the KSNP to show the usefulness of it. The application results of the developed guideline show that more than 50% of the items to be improved for pre-accident human errors of auxiliary feed water system are resolved. The guideline for modeling pre-accident human errors developed in this study can be used for other NPPs as well as the KSNP. It is expected that both use of the detailed procedure, to be developed in the future, for the quantification of pre-accident human errors and the guideline developed in this study will greatly enhance the PSA quality in the HRA of pre-accident human errors.

  15. Development of an accident sequence precursor methodology and its application to significant accident precursors

    Energy Technology Data Exchange (ETDEWEB)

    Jang, Seung Hyun; Park, Sung Hyun; Jae, Moo Sung [Dept. of of Nuclear Engineering, Hanyang University, Seoul (Korea, Republic of)

    2017-03-15

    The systematic management of plant risk is crucial for enhancing the safety of nuclear power plants and for designing new nuclear power plants. Accident sequence precursor (ASP) analysis may be able to provide risk significance of operational experience by using probabilistic risk assessment to evaluate an operational event quantitatively in terms of its impact on core damage. In this study, an ASP methodology for two operation mode, full power and low power/shutdown operation, has been developed and applied to significant accident precursors that may occur during the operation of nuclear power plants. Two operational events, loss of feedwater and steam generator tube rupture, are identified as ASPs. Therefore, the ASP methodology developed in this study may contribute to identifying plant risk significance as well as to enhancing the safety of nuclear power plants by applying this methodology systematically.

  16. The role of work stress in the development of accidents

    NARCIS (Netherlands)

    Vaas, F.

    2003-01-01

    At December 2nd, 2003 the Third International Tripod Symposium: Loss of Control was took place. At the symposium a lecture was held by Dr. F. Vaas of TNO Work and Employment about the role of work stress in the development of accidents. At TNO Work and Employment a management tool was developed to

  17. Dust resuspension and transport modeling for loss of vacuum accidents

    International Nuclear Information System (INIS)

    Humrickhouse, P.W.; Corradini, M.L.; Sharpe, J.P.

    2007-01-01

    Plasma surface interactions in tokamaks are known to create significant quantities of dust, which settles onto surfaces and accumulates in the vacuum vessel. In ITER, a loss of vacuum accident may result in the release of dust which will be radioactive and/or toxic, and provides increased surface area for chemical reactions or dust explosion. A new method of analysis has been developed for modeling dust resuspension and transport in loss of vacuum accidents. The aerosol dynamic equation is solved via the user defined scalar (UDS) capability in the commercial CFD code Fluent. Fluent solves up to 50 generic transport equations for user defined scalars, and allows customization of terms in these equations through user defined functions (UDF). This allows calculation of diffusion coefficients based on local flow properties, inclusion of body forces such as gravity and thermophoresis in the convection term, and user defined source terms. The code accurately reproduces analytical solutions for aerosol deposition in simple laminar flows with diffusion and gravitational settling. Models for dust resuspension are evaluated, and code results are compared to available resuspension data, including data from the Toroidal Dust Mobilization Experiment (TDMX) at the Idaho National Laboratory. Extension to polydisperse aerosols and inclusion of coagulation effects is also discussed. (orig.)

  18. Accident tolerant fuel cladding development: Promise, status, and challenges

    Science.gov (United States)

    Terrani, Kurt A.

    2018-04-01

    The motivation for transitioning away from zirconium-based fuel cladding in light water reactors to significantly more oxidation-resistant materials, thereby enhancing safety margins during severe accidents, is laid out. A review of the development status for three accident tolerant fuel cladding technologies, namely coated zirconium-based cladding, ferritic alumina-forming alloy cladding, and silicon carbide fiber-reinforced silicon carbide matrix composite cladding, is offered. Technical challenges and data gaps for each of these cladding technologies are highlighted. Full development towards commercial deployment of these technologies is identified as a high priority for the nuclear industry.

  19. A systematic process for developing and assessing accident management plans

    International Nuclear Information System (INIS)

    Hanson, D.J.; Blackman, H.S.; Meyer, O.R.; Ward, L.W.

    1991-04-01

    This document describes a four-phase approach for developing criteria recommended for use in assessing the adequacy of nuclear power plant accident management plans. Two phases of the approach have been completed and provide a prototype process that could be used to develop an accident management plan. Based on this process, a preliminary set of assessment criteria are derived. These preliminary criteria will be refined and improved when the remaining steps of the approach are completed, that is, after the prototype process is validated through application. 9 refs., 10 figs., 7 tabs

  20. Development of source term PIRT of Fukushima Daiichi NPPs accident

    International Nuclear Information System (INIS)

    Suehiro, S.; Okamoto, K.

    2017-01-01

    The severe accident evaluation committee of AESJ (Atomic Energy Society of Japan) developed the thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and the source term PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aimed to explore the debris distribution and the current condition in the NPPs with high accuracy and to extract higher priority from the aspect of the sophistication of the analytical technology to predict the severe accident phenomena by the code. The source term PIRT was divided into 3 phases for the time domain and 9 categories for the spatial domain. The 68 phenomena were extracted and the importance from viewpoint of the source term was ranked through brainstorming and discussion. This paper describes the developed source term PIRT list and summarized the high ranked phenomena in each phase. (author)

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

    International Nuclear Information System (INIS)

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

    2014-01-01

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

  2. Development Status of Accident Tolerant Fuel Cladding for LWRs

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Hyun-Gil; Kim, Il-Hyun; Jung, Yang-Il; Park, Dong-Jun; Park, Jung-Hwan; Yang, Jae-Ho; Koo, Yang-Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2016-10-15

    Hydrogen explosions and the release of radionuclides are caused by severe damage of current nuclear fuels, which are composed of fuel pellets and fuel cladding, during an accident. To reduce the damage to the public, the fuels have to enhance their integrity under an accident environment. Enhanced accident tolerance fuels (ATFs) can tolerate a loss of active cooling in the reactor core for a considerably longer time period during design-basis and beyond design-basis events while maintaining or improving the fuel performance during normal operations as well as operational transients, in comparison with the current UO{sub 2}-Zr alloy system used in the LWR. Surface modified Zr cladding as a new concept was suggested to apply an enhanced ATF cladding. The aim of the partial ODS treatment is to increase the high-temperature strength to suppress the ballooning/rupture behavior of fuel cladding during an accident event. The target of the surface coating is to increase the corrosion resistance during normal operation and increase the oxidation resistance during an accident event. The partial ODS treatment of Zircaloy-4 cladding can be produced using a laser beam scanning method with Y2O3 powder, and the surface Cr-alloy and Cr/FeCrAl coating on Zircaloy-4 cladding can be obtained after the development of 3D laser coating and arc ion plating technologies.

  3. Models of fuel masses transition during second stage of the accident on Chernobyl NPP

    International Nuclear Information System (INIS)

    Tarapon, A.

    2002-01-01

    In ISPE NASU of Ukraine are developed mathematical models and software, which allow to research the processes of fuel masses transition during the accident at ChNPP. We found out, that the main reason of accident on ChNPP is the happening in the reactor of crisis of heat exchange of the second sort, instead of the effect positive output of reactivity from displacers of rods of system of emergency protection, as is accepted in official version

  4. Model for calculation of concentration and load on behalf of accidents with radioactive materials

    International Nuclear Information System (INIS)

    Janssen, L.A.M.; Heugten, W.H.H. van

    1987-04-01

    In the project 'Information- and calculation-system for disaster combatment', by order of the Dutch government, a demonstration model has been developed for a diagnosis system for accidents. In this demonstration a model is used to calculate the concentration- and dose-distributions caused by incidental emissions of limited time. This model is described in this report. 4 refs.; 2 figs.; 3 tabs

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

    International Nuclear Information System (INIS)

    Kim, Heui Dong; Cho, Sung Won; Park, Jong Hwa; Hong, Sung Wan; Yoo, Dong Han; Hwang, Moon Kyoo; Noh, Kee Man; Song, Yong Man

    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 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)

  6. Fission-gas bubble modeling for LMFBR accidents

    International Nuclear Information System (INIS)

    Ostensen, R.W.

    1977-01-01

    The behavior of fission-gas bubbles in unrestructured oxide fuel can have a dominant effect on the course of a core disruptive accident in an LMFBR. The paper describes a simplified model of bubble behavior and presents results of that model in analyzing the relevant physical assumptions and predicting gas behavior in molten fuel

  7. Hull loss accident model for narrow body commercial aircraft

    Directory of Open Access Journals (Sweden)

    Somchanok Tiabtiamrat

    2010-10-01

    Full Text Available Accidents with narrow body aircraft were statistically evaluated covering six families of commercial aircraft includingBoeing B737, Airbus A320, McDonnell Douglas MD80, Tupolev TU134/TU154 and Antonov AN124. A risk indicator for eachflight phase was developed based on motion characteristics, duration time, and the presence of adverse weather conditions.The estimated risk levels based on these risk indicators then developed from the risk indicator. Regression analysis indicatedvery good agreement between the estimated risk level and the accident ratio of hull loss cases per number of delivered aircraft.The effect of time on the hull loss accident ratio per delivered aircraft was assessed for B737, A320 and MD80. Equationsrepresenting the effect of time on hull loss accident ratio per delivered aircraft were proposed for B737, A320, and MD80,while average values of hull loss accident ratio per delivered aircraft were found for TU134, TU154, and AN 124. Accidentprobability equations were then developed for each family of aircraft that the probability of an aircraft in a hull loss accidentcould be estimated for any aircraft family, flight phase, presence of adverse weather factor, hour of day, day of week, monthof year, pilot age, and pilot flight hour experience. A simplified relationship between estimated hull loss accident probabilityand unsafe acts by human was proposed. Numerical investigation of the relationship between unsafe acts by human andfatality ratio suggested that the fatality ratio in hull loss accident was dominated primarily by the flight phase media.

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

    International Nuclear Information System (INIS)

    Tusheva, Polina

    2012-01-01

    effectiveness of the procedures strongly depends on the ability of the passive safety systems to inject as much water as possible into the reactor coolant system. The results on the early in-vessel phase have shown potentially delayed RPV failure by depressurization of the primary side, as slowing the core damage gives more time and different possibilities for operator interventions to recover systems and to mitigate or terminate the accident. The ANSYS model for the description of the molten pool behaviour in the RPV lower plenum has been extended by a model considering a stratified molten pool configuration. Two different pool configurations were analysed: homogeneous and segregated. The possible failure modes of the RPV and the time to failure were investigated to assess the possible loadings on the containment. The main treated issues are: the temperature field within the corium pool and the RPV and the structure-mechanical behaviour of the vessel wall. The results of the ASTEC calculations of the melt pool configuration were applied as initial conditions for the ANSYS simulations, allowing a more detailed and more accurate modelling of the thermal and mechanical behaviour of the core melt and the RPV wall. Moreover, for the late in-vessel phase, retention of the corium in the RPV was investigated presuming external cooling of the vessel wall as mitigative severe accident management measure. The study was based on the finite element computer code ANSYS. The highest thermomechanical loads are observed in the transition zone between the elliptical and the vertical vessel wall for homogeneous pool and in the vertical part of the vessel wall, which is in contact with the molten metal in case of sub-oxidized pool. Assuming external flooding will retain the corium within the RPV. Without flooding, the vessel wall will fail, as the necessary temperature for a balanced heat release from the external surface via radiation is near to or above the melting point of the steel.

  9. Application of the accident consequences model of the German risk study to assessments of accident risks in different types of nuclear power plants

    International Nuclear Information System (INIS)

    Ehrhardt, J.; Bayer, A.

    1982-01-01

    Within the scope of the 'German Risk Study for Nuclear Power Plants' (Phase A) the accident consequence model UFOMOD was developed in the Karlsruhe Nuclear Research Center. This model originally developed for pressurized water reactors has now been extended in order to obtain results about accidental releases of activity from fast breeder and high-temperature reactors, too. (RW) [de

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

    International Nuclear Information System (INIS)

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

    2005-01-01

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

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

    International Nuclear Information System (INIS)

    Carmen Carnero, Maria; Jose Pedregal, Diego

    2010-01-01

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

  12. Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents

    International Nuclear Information System (INIS)

    Chang, Y.H.J.; Mosleh, A.

    2007-01-01

    This is the last in a series of five papers that discuss the Information Decision and Action in Crew (IDAC) context for human reliability analysis (HRA) and example application. The model is developed to probabilistically predict the responses of the control room operating crew in nuclear power plants during an accident, for use in probabilistic risk assessments (PRA). The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper describes a dynamic PRA computer simulation program, accident dynamics simulator (ADS), developed in part to implement the IDAC model. This paper also provides a detailed example of implementing a simpler version of IDAC, compared with the IDAC model discussed in the first four papers of this series, to demonstrate the practicality of integrating a detailed cognitive HRA model within a dynamic PRA framework

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-05-15

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

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

    International Nuclear Information System (INIS)

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

    2002-03-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-07-01

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

  16. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    International Nuclear Information System (INIS)

    Konoval, A.V.; Kalvand, Ali.; Kazachkov, I.V.

    2013-01-01

    The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS) against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks' distributions considered and the results of their work in the corium cooling pool are compared to the data of other PPS's conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents

  17. Modeling of the corium cooling and loading factor analysis for containment during severe accidents

    Directory of Open Access Journals (Sweden)

    O. V. Konoval

    2013-09-01

    Full Text Available The paper is devoted to the development and study of the mathematical model for corium melt interaction with low-temperature melting blocks in the passive protection systems (PPS against severe accidents at the NPP, and learning the peculiarities of construction and operation of the PPS. The configurations of cooling blocks’ distributions considered and the results of their work in the corium cooling pool are compared to the data of oth-er PPS’s conceptions. The conclusion is made that the models developed and the results obtained may be useful for constructing the PPS against severe accidents.

  18. Reactor accident calculation models in use in the Nordic countries

    International Nuclear Information System (INIS)

    Tveten, U.

    1984-01-01

    The report relates to a subproject under a Nordic project called ''Large reactor accidents - consequences and mitigating actions''. In the first part of the report short descriptions of the various models are given. A systematic list by subject is then given. In the main body of the report chapter and subchapter headings are by subject. (Auth.)

  19. Accident Analysis Methods and Models — a Systematic Literature Review

    NARCIS (Netherlands)

    Wienen, Hans Christian Augustijn; Bukhsh, Faiza Allah; Vriezekolk, E.; Wieringa, Roelf J.

    2017-01-01

    As part of our co-operation with the Telecommunication Agency of the Netherlands, we want to formulate an accident analysis method and model for use in incidents in telecommunications that cause service unavailability. In order to not re-invent the wheel, we wanted to first get an overview of all

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-07-01

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

  1. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    International Nuclear Information System (INIS)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul

    2015-01-01

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree

  2. Restructuring of an Event Tree for a Loss of Coolant Accident in a PSA model

    Energy Technology Data Exchange (ETDEWEB)

    Lim, Ho-Gon; Han, Sang-Hoon; Park, Jin-Hee; Jang, Seong-Chul [KAERI, Daejeon (Korea, Republic of)

    2015-05-15

    Conventional risk model using PSA (probabilistic Safety Assessment) for a NPP considers two types of accident initiators for internal events, LOCA (Loss of Coolant Accident) and transient event such as Loss of electric power, Loss of cooling, and so on. Traditionally, a LOCA is divided into three initiating event (IE) categories depending on the break size, small, medium, and large LOCA. In each IE group, safety functions or systems modeled in the accident sequences are considered to be applicable regardless of the break size. However, since the safety system or functions are not designed based on a break size, there exist lots of mismatch between safety system/function and an IE, which may make the risk model conservative or in some case optimistic. Present paper proposes new methodology for accident sequence analysis for LOCA. We suggest an integrated single ET construction for LOCA by incorporating a safety system/function and its applicable break spectrum into the ET. Integrated accident sequence analysis in terms of ET for LOCA was proposed in the present paper. Safety function/system can be properly assigned if its applicable range is given by break set point. Also, using simple Boolean algebra with the subset of the break spectrum, final accident sequences are expressed properly in terms of the Boolean multiplication, the occurrence frequency and the success/failure of safety system. The accident sequence results show that the accident sequence is described more detailed compared with the conventional results. Unfortunately, the quantitative results in terms of MCS (minimal Cut-Set) was not given because system fault tree was not constructed for this analysis and the break set points for all 7 point were not given as a specified numerical quantity. Further study may be needed to fix the break set point and to develop system fault tree.

  3. Having a New Pair of Glassess : Applying Systemic Accident Models on Road Safety

    OpenAIRE

    Huang, Yu-Hsing

    2007-01-01

    The main purpose of the thesis is to discuss the accident models which underlie accident prevention in general and road safety in particular, and the consequences of relying on a particular model have for actual preventive work. The discussion centres on two main topics. The first topic is whether the underlying accident model, or paradigm, of traditional road safety should be exchanged for a more complex accident model, and if so, which model(s) are appropriate. From a discussion of current ...

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

    International Nuclear Information System (INIS)

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

    1978-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1991-12-15

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

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

    International Nuclear Information System (INIS)

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

    1991-12-01

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

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

    International Nuclear Information System (INIS)

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

    2004-01-01

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

  8. Development status of Severe Accident Analysis Code SAMPSON

    International Nuclear Information System (INIS)

    Iwashita, Tsuyoshi; Ujita, Hiroshi

    2000-01-01

    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)

  9. Object-Oriented Bayesian Networks (OOBN) for Aviation Accident Modeling and Technology Portfolio Impact Assessment

    Science.gov (United States)

    Shih, Ann T.; Ancel, Ersin; Jones, Sharon M.

    2012-01-01

    The concern for reducing aviation safety risk is rising as the National Airspace System in the United States transforms to the Next Generation Air Transportation System (NextGen). The NASA Aviation Safety Program is committed to developing an effective aviation safety technology portfolio to meet the challenges of this transformation and to mitigate relevant safety risks. The paper focuses on the reasoning of selecting Object-Oriented Bayesian Networks (OOBN) as the technique and commercial software for the accident modeling and portfolio assessment. To illustrate the benefits of OOBN in a large and complex aviation accident model, the in-flight Loss-of-Control Accident Framework (LOCAF) constructed as an influence diagram is presented. An OOBN approach not only simplifies construction and maintenance of complex causal networks for the modelers, but also offers a well-organized hierarchical network that is easier for decision makers to exploit the model examining the effectiveness of risk mitigation strategies through technology insertions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-12-15

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

  11. Chernobyl and Fukushima nuclear accidents: what has changed in the use of atmospheric dispersion modeling?

    International Nuclear Information System (INIS)

    Benamrane, Y.; Wybo, J.-L.; Armand, P.

    2013-01-01

    The threat of a major accidental or deliberate event that would lead to hazardous materials emission in the atmosphere is a great cause of concern to societies. This is due to the potential large scale of casualties and damages that could result from the release of explosive, flammable or toxic gases from industrial plants or transport accidents, radioactive material from nuclear power plants (NPPs), and chemical, biological, radiological or nuclear (CBRN) terrorist attacks. In order to respond efficiently to such events, emergency services and authorities resort to appropriate planning and organizational patterns. This paper focuses on the use of atmospheric dispersion modeling (ADM) as a support tool for emergency planning and response, to assess the propagation of the hazardous cloud and thereby, take adequate counter measures. This paper intends to illustrate the noticeable evolution in the operational use of ADM tools over 25 y and especially in emergency situations. This study is based on data available in scientific publications and exemplified using the two most severe nuclear accidents: Chernobyl (1986) and Fukushima (2011). It appears that during the Chernobyl accident, ADM were used few days after the beginning of the accident mainly in a diagnosis approach trying to reconstruct what happened, whereas 25 y later, ADM was also used during the first days and weeks of the Fukushima accident to anticipate the potentially threatened areas. We argue that the recent developments in ADM tools play an increasing role in emergencies and crises management, by supporting stakeholders in anticipating, monitoring and assessing post-event damages. However, despite technological evolutions, its prognostic and diagnostic use in emergency situations still arise many issues. -- Highlights: • Study of atmospheric dispersion modeling use during nuclear accidents. • ADM tools were mainly used in a diagnosis approach during Chernobyl accident. • ADM tools were also used

  12. Study on Developments in Accident Investigation Methods: A Survey of the 'State-of-the-Art'

    International Nuclear Information System (INIS)

    Hollnagel, Erik; Speziali, Josephine

    2008-01-01

    The objective of this project was to survey the main accident investigation methods that have been developed since the early or mid-1990s. The motivation was the increasing frequency of accidents that defy explanations in simple terms, for instance cause-effect chains or 'human error'. Whereas the complexity of socio-technical systems is steadily growing across all industrial domains, including nuclear power production, accident investigation methods are only updated when their inability to account for novel types of accidents and incidents becomes inescapable. Accident investigation methods therefore typically lag behind the socio-technological developments by 20 years or more. The project first compiled a set of methods from the recognised scientific literature and in major major research and development programs, excluding methods limited to risk assessment, technological malfunctions, human reliability, and safety management methods. An initial set of 21 methods was further reduced to seven by retaining only prima facie accident investigation methods and avoiding overlapping or highly similar methods. The second step was to develop a set of criteria used to characterise the methods. The starting point was Perrow's description of normal accidents in socio-technical systems, which used the dimensions of coupling, going from loose to tight, and interactions, going from linear to complex. For practical reasons, the second dimension was changed to that of tractability or how easy it is to describe the system, where the sub-criteria are the level of detail, the availability of an articulated model, and the system dynamics. On this basis the seven selected methods were characterised in terms of the systems - or conditions - they could account for, leading to the following four groups: methods suitable for systems that are loosely coupled and tractable, methods suitable for systems that are tightly coupled and tractable, methods suitable for systems that are loosely

  13. Development and verification of a three-dimensional core model for WWR type reactors and its coupling with the accident code ATHLET. Final report

    International Nuclear Information System (INIS)

    Grundmann, U.; Lucas, D.; Mittag, S.; Rohde, U.

    1995-04-01

    The main goal of the project was the coupling of the 3D core model DYN3D for Russian VVER-type reactors, which has been developed in the RCR, with the thermohydraulic code ATHLET. The coupling has been realized on two basically different ways: - The implementation of only the neutron kinetics model of DYN3D into ATHLET (internal coupling), - the connection of the complete DYN3D core model including neutron kinetics, thermohydraulics and fuel rod model via data interfaces at the core top and bottom (external coupling). For the test of the coupling, comparative calculations between internal and external coupling versions have been carried out for a LOCA and a reactivity transient. Complementary goals of the project were: - The development of a DYN3D version for burn-up calculations, - the verification of DYN3D on benchmark tasks and experimental data on fuel rod behaviour, - a study on the extension of the neutron-physical data base. The project contributed to the development of advanced tools for the safety analysis of VVER-type reactors. Future work is aimed to the verification of the coupled code complex DYN3D-ATHLET. (orig.) [de

  14. Use of accident experience in developing criteria for teleoperator equipment

    International Nuclear Information System (INIS)

    Vallario, E.J.; Selby, J.M.

    1985-10-01

    The 1961 SL-1 reactor accident in Idaho and the Recuplex accident at Hanford are reviewed to identify problems common to emergency situations, lessons learned from accidents, criteria for emergency equipment, and recommendations for using robotics to solve problems during emergencies. Teleoperator equipment could be used to assess the extent of the damage and the condition of the reactor, retrieve dosimeters, evacuate and treat accident victims, clean up debris and decontaminate accident areas. 2 refs., 9 figs

  15. Illustration interface of accident progression in PWR by quick inference based on multilevel flow models

    International Nuclear Information System (INIS)

    Yoshikawa, H.; Ouyang, J.; Niwa, Y.

    2006-01-01

    In this paper, a new accident inference method is proposed by using a goal and function oriented modeling method called Multilevel Flow Model focusing on explaining the causal-consequence relations and the objective of automatic action in the accident of nuclear power plant. Users can easily grasp how the various plant parameters will behave and how the various safety facilities will be activated sequentially to cope with the accident until the nuclear power plants are settled into safety state, i.e., shutdown state. The applicability of the developed method was validated by the conduction of internet-based 'view' experiment to the voluntary respondents, and in the future, further elaboration of interface design and the further introduction of instruction contents will be developed to make it become the usable CAI system. (authors)

  16. A Review of Accident Modelling Approaches for Complex Critical Sociotechnical Systems

    National Research Council Canada - National Science Library

    Qureshi, Zahid H

    2008-01-01

    .... This report provides a review of key traditional accident modelling approaches and their limitations, and describes new system-theoretic approaches to the modelling and analysis of accidents in safety-critical systems...

  17. Impacts of the Fukushima Daiichi Accident on Nuclear Development Policies

    International Nuclear Information System (INIS)

    Vance, Robert; Henderson, David; ); Moore, Laurie

    2017-01-01

    The Fukushima Daiichi nuclear power plant accident has had an impact on the development of nuclear power around the world. While the accident was followed by thorough technical assessments of the safety of all operating nuclear power plants, and a general increase in safety requirements has been observed worldwide, national policy responses have been more varied. These responses have ranged from countries phasing out or accelerating decisions to phase out nuclear energy to countries reducing their reliance on nuclear power or on the contrary continuing to pursue or expand their nuclear power programs. This study examines changes to policies, and plans and attempts to distinguish the impact of the Fukushima Daiichi accident from other factors that have affected policy-making in relation to nuclear energy, in particular electricity market economics, financing challenges and competition from other sources (gas, coal and renewables). It also examines changes over time to long-term, quantitative country projections, which reveal interesting trends on the possible role of nuclear energy in future energy systems. (authors)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-09-01

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

  19. A flammability and combustion model for integrated accident analysis

    International Nuclear Information System (INIS)

    Plys, M.G.; Astleford, R.D.; Epstein, M.

    1988-01-01

    A model for flammability characteristics and combustion of hydrogen and carbon monoxide mixtures is presented for application to severe accident analysis of Advanced Light Water Reactors (ALWR's). Flammability of general mixtures for thermodynamic conditions anticipated during a severe accident is quantified with a new correlation technique applied to data for several fuel and inertant mixtures and using accepted methods for combining these data. Combustion behavior is quantified by a mechanistic model consisting of a continuity and momentum balance for the burned gases, and considering an uncertainty parameter to match the idealized process to experiment. Benchmarks against experiment demonstrate the validity of this approach for a single recommended value of the flame flux multiplier parameter. The models presented here are equally applicable to analysis of current LWR's. 21 refs., 16 figs., 6 tabs

  20. Boundary effects on car accidents in a cellular automaton model

    International Nuclear Information System (INIS)

    Yang Xianqing; Ma Yuqiang; Zhao Yuemin

    2004-01-01

    In this paper we numerically study the probability P ac of occurrence of car accidents in the Nagel-Schreckenberg (NS) model with open boundary condition. In the deterministic NS model, numerical results show that there exists a critical value of extinction rate β above which no car accidents occur, and below which the probability P ac is independent of the speed limit v max and the injection rate α, but only determined by the extinction rate β. In the non-deterministic NS model, the probability P ac is a non-monotonic function of β in the region of low β value, while it is independent of β in the region of high β value. The stochastic braking not only reduces the occurrence of car accidents, but splits degenerate effects of v max on the probability P ac . Theoretical analyses give an agreement with numerical results in the deterministic NS model and in the non-deterministic NS model with v max = 1 in the case of low β value region. Qualitative differences between open and periodic systems in the relations of P ac to the bulk density ρ imply that various correlations may exist between the two systems

  1. Development of likelihood estimation method for criticality accidents of mixed oxide fuel fabrication facilities

    International Nuclear Information System (INIS)

    Tamaki, Hitoshi; Yoshida, Kazuo; Kimoto, Tatsuya; Hamaguchi, Yoshikane

    2010-01-01

    A criticality accident in a MOX fuel fabrication facility may occur depending on several parameters, such as mass inventory and plutonium enrichment. MOX handling units in the facility are designed and operated based on the double contingency principle to prevent criticality accidents. Control failures of at least two parameters are needed for the occurrence of criticality accident. To evaluate the probability of such control failures, the criticality conditions of each parameter for a specific handling unit are necessary for accident scenario analysis to be clarified quantitatively with a criticality analysis computer code. In addition to this issue, a computer-based control system for mass inventory is planned to be installed into MOX handling equipment in a commercial MOX fuel fabrication plant. The reliability analysis is another important issue in evaluating the likelihood of control failure caused by software malfunction. A likelihood estimation method for criticality accident has been developed with these issues been taken into consideration. In this paper, an example of analysis with the proposed method and the applicability of the method are also shown through a trial application to a model MOX fabrication facility. (author)

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

    International Nuclear Information System (INIS)

    2015-11-01

    The demands on nuclear fuel have recently been increasing, and include transient regimes, higher discharge burnup and longer fuel cycles. This has resulted in an increase of loads on fuel and core internals. In order to satisfy these demands while ensuring compliance with safety criteria, new national and international programmes have been launched and advanced modelling codes are being developed. The Fukushima Daiichi accident has particularly demonstrated the need for adequate analysis of all aspects of fuel performance to prevent a failure and also to predict fuel behaviour were an accident to occur.This publication presents the Proceedings of the Technical Meeting on Modelling of Water Cooled Fuel Including Design Basis and Severe Accidents, which was hosted by the Nuclear Power Institute of China (NPIC) in Chengdu, China, following the recommendation made in 2013 at the IAEA Technical Working Group on Fuel Performance and Technology. This recommendation was in agreement with IAEA mid-term initiatives, linked to the post-Fukushima IAEA Nuclear Safety Action Plan, as well as the forthcoming Coordinated Research Project (CRP) on Fuel Modelling in Accident Conditions. At the technical meeting in Chengdu, major areas and physical phenomena, as well as types of code and experiment to be studied and used in the CRP, were discussed. The technical meeting provided a forum for international experts to review the state of the art of code development for modelling fuel performance of nuclear fuel for water cooled reactors with regard to steady state and transient conditions, and for design basis and early phases of severe accidents, including experimental support for code validation. A round table discussion focused on the needs and perspectives on fuel modelling in accident conditions. This meeting was the ninth in a series of IAEA meetings, which reflects Member States’ continuing interest in nuclear fuel issues. The previous meetings were held in 1980 (jointly with

  3. Phenomenological and mechanistic modeling of melt-structure-water interactions in a light water reactor severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Bui, V.A

    1998-10-01

    The objective of this work is to address the modeling of the thermal hydrodynamic phenomena and interactions occurring during the progression of reactor severe accidents. Integrated phenomenological models are developed to describe the accident scenarios, which consist of many processes, while mechanistic modeling, including direct numerical simulation, is carried out to describe separate effects and selected physical phenomena of particular importance 88 refs, 54 figs, 7 tabs

  4. Uncertainty and sensitivity analysis in reactivity-initiated accident fuel modeling: synthesis of organisation for economic co-operation and development (OECD/nuclear energy agency (NEA benchmark on reactivity-initiated accident codes phase-II

    Directory of Open Access Journals (Sweden)

    Olivier Marchand

    2018-03-01

    Full Text Available In the framework of OECD/NEA Working Group on Fuel Safety, a RIA fuel-rod-code Benchmark Phase I was organized in 2010–2013. It consisted of four experiments on highly irradiated fuel rodlets tested under different experimental conditions. This benchmark revealed the need to better understand the basic models incorporated in each code for realistic simulation of the complicated integral RIA tests with high burnup fuel rods. A second phase of the benchmark (Phase II was thus launched early in 2014, which has been organized in two complementary activities: (1 comparison of the results of different simulations on simplified cases in order to provide additional bases for understanding the differences in modelling of the concerned phenomena; (2 assessment of the uncertainty of the results. The present paper provides a summary and conclusions of the second activity of the Benchmark Phase II, which is based on the input uncertainty propagation methodology. The main conclusion is that uncertainties cannot fully explain the difference between the code predictions. Finally, based on the RIA benchmark Phase-I and Phase-II conclusions, some recommendations are made. Keywords: RIA, Codes Benchmarking, Fuel Modelling, OECD

  5. Cost-effectiveness analysis of countermeasures using accident consequence assessment models

    International Nuclear Information System (INIS)

    Alonso, A.; Gallego, E.

    1987-01-01

    In the event of a large release of radionuclides from a nuclear power plant, protective actions for the population potentially affected must be implemented. Cost-effectiveness analysis will be useful to define the countermeasures and the criteria needed to implement them. This paper shows the application of Accident Consequence Assessment (ACA) models to cost-effectiveness analysis of emergency and long-term countermeasures, making use of the different relationships between dose, contamination levels, affected areas and population distribution, included in such a model. The procedure is illustrated with the new Melcor Accident Consequence Code System (MACCS 1.3), developed at Sandia National Laboratories (USA), for a fixed accident scenario. Different alternative actions are evaluated with regard to their radiological and economical impact, searching for an 'optimum' strategy. (author)

  6. Development of an Accident Reproduction Simulator System Using a Hemodialysis Extracorporeal Circulation System

    OpenAIRE

    Nishite, Yoshiaki; Takesawa, Shingo

    2016-01-01

    Background Accidents that occur during dialysis treatment are notified to the medical staff via alarms raised by the dialysis apparatus. Similar to such real accidents, apparatus activation or accidents can be reproduced by simulating a treatment situation. An alarm that corresponds to such accidents can be utilized in the simulation model. Objectives The aim of this study was to create an extracorporeal circulation system (herein...

  7. Accident consequence assessments with different atmospheric dispersion models

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1989-11-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straight-line Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different dispersion models on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been performed. The study showed that there are trajectory models available which can be applied in ACAs and that they provide more realistic results of ACAs than straight-line Gaussian models. This led to a completely novel concept of atmospheric dispersion modelling in which two different distance ranges of validity are distinguished: the near range of some ten kilometres distance and the adjacent far range which are assigned to respective trajectory models. (orig.) [de

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-15

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

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

    International Nuclear Information System (INIS)

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

    1992-01-01

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

  10. Development of an accident diagnosis system using a dynamic neural network for nuclear power plants

    International Nuclear Information System (INIS)

    Lee, Seung Jun; Kim, Jong Hyun; Seong, Poong Hyun

    2004-01-01

    In this work, an accident diagnosis system using the dynamic neural network is developed. In order to help the plant operators to quickly identify the problem, perform diagnosis and initiate recovery actions ensuring the safety of the plant, many operator support system and accident diagnosis systems have been developed. Neural networks have been recognized as a good method to implement an accident diagnosis system. However, conventional accident diagnosis systems that used neural networks did not consider a time factor sufficiently. If the neural network could be trained according to time, it is possible to perform more efficient and detailed accidents analysis. Therefore, this work suggests a dynamic neural network which has different features from existing dynamic neural networks. And a simple accident diagnosis system is implemented in order to validate the dynamic neural network. After training of the prototype, several accident diagnoses were performed. The results show that the prototype can detect the accidents correctly with good performances

  11. Statistical aspects of carbon fiber risk assessment modeling. [fire accidents involving aircraft

    Science.gov (United States)

    Gross, D.; Miller, D. R.; Soland, R. M.

    1980-01-01

    The probabilistic and statistical aspects of the carbon fiber risk assessment modeling of fire accidents involving commercial aircraft are examined. Three major sources of uncertainty in the modeling effort are identified. These are: (1) imprecise knowledge in establishing the model; (2) parameter estimation; and (3)Monte Carlo sampling error. All three sources of uncertainty are treated and statistical procedures are utilized and/or developed to control them wherever possible.

  12. Development of an Accident Diagnostic Scheme Using Artificial Intelligence Techniques (I)

    Energy Technology Data Exchange (ETDEWEB)

    Na, M. G.; Lee, S. H.; Kim, D. S.; No, Y. G.; Lee, S. W. [Chosun University, Gwangju (Korea, Republic of); Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-06-15

    As a means to effectively manage the severe nuclear accidents, it is important to identify and diagnose the accident initiating events during an initial short time interval after the accidents by observing the major controlling parameters. Main objective of this study is to develop the diagnostic approach for the accurate prediction of accident initiating events using artificial intelligence techniques. For this, first, a variety of artificial intelligence techniques such as Finn, Gmbh, and Sm were examined through this study. Among them, Sc and Gmbh model were assessed as a useful approach to predict the break location and the break size of Local. In order to verify the proposed algorithm, the 111 accident simulation data (based on Map) were applied to train the Sc and Gmbh models, and the test data was used to independently verify whether or not the SVC and GMDH models work well. The analysis of the maximum errors and RMS errors, and the performance of the GMDH according to the existence of measurement errors and SIS actuation showed that the proposed SVC and GMDH models can accurately classify the break locations and accurately predict the break size. As the time-integrated signals were used for inputs into the GMDH model within a period of 60 second after a reactor scram, the actuation of the safety systems such as safety injection system (SIS), auxiliary feed water system, and containment spray system, were not considered in this study. It is because the initial 60 second time-integrated signals were used and the safety systems usually start to actuate after a more than 60 second time delay after the reactor scram

  13. Development of an Accident Diagnostic Scheme Using Artificial Intelligence Techniques (I)

    International Nuclear Information System (INIS)

    Na, M. G.; Lee, S. H.; Kim, D. S.; No, Y. G.; Lee, S. W.; Ahn, K. I.

    2010-06-01

    As a means to effectively manage the severe nuclear accidents, it is important to identify and diagnose the accident initiating events during an initial short time interval after the accidents by observing the major controlling parameters. Main objective of this study is to develop the diagnostic approach for the accurate prediction of accident initiating events using artificial intelligence techniques. For this, first, a variety of artificial intelligence techniques such as Finn, Gmbh, and Sm were examined through this study. Among them, Sc and Gmbh model were assessed as a useful approach to predict the break location and the break size of Local. In order to verify the proposed algorithm, the 111 accident simulation data (based on Map) were applied to train the Sc and Gmbh models, and the test data was used to independently verify whether or not the SVC and GMDH models work well. The analysis of the maximum errors and RMS errors, and the performance of the GMDH according to the existence of measurement errors and SIS actuation showed that the proposed SVC and GMDH models can accurately classify the break locations and accurately predict the break size. As the time-integrated signals were used for inputs into the GMDH model within a period of 60 second after a reactor scram, the actuation of the safety systems such as safety injection system (SIS), auxiliary feed water system, and containment spray system, were not considered in this study. It is because the initial 60 second time-integrated signals were used and the safety systems usually start to actuate after a more than 60 second time delay after the reactor scram

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

    International Nuclear Information System (INIS)

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

    2001-01-01

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

  15. Processing Expert Judgements in Accident Consequence Modelling (invited paper)

    International Nuclear Information System (INIS)

    Kraan, B.C.P.; Cooke, R.M.

    2000-01-01

    In performing uncertainty analysis a distribution on the code input parameters is required. The construction of the distribution on the code input parameters for the joint CEC/USNRC Accident Consequence Code Uncertainty Analysis using Expert Judgement is discussed. An example from the food chain module is used to illustrate the construction. Different mathematical techniques have been developed to transform the expert judgements into the required format. Finally, the effect of taking account of correlations in performing uncertainty analysis is investigated. (author)

  16. Derived intervention levels in early stage of nuclear accident development

    Energy Technology Data Exchange (ETDEWEB)

    Vladar, M; Fojtik, M [Inst. of Preventive and Clinical Medicine, Bratislava (Slovakia); Stubna, M [Research Inst. of Nuclear Power Plants, Bohunuce (Slovakia)

    1996-12-31

    In this paper the measures for protection of health and property of public in the case of nuclear accident are discussed. They are based on optimal application of so called intervention levels. The actual flow of decision depends on: (1) prognosis of mathematical modelling of possible course of nuclear accident, and (2) results of monitoring of radiation situation.The aim of this contribution was to analyze their mutual cooperation and to suggest such procedure of monitoring or radiation situation which could be used for suggestion of protective measures. In this contribution the zones of protection planning in the accident place surrounding for the urgent measures were specified : (1) regulation of free movement of persons; (2) sheltering; (3) iodine prophylaxis; (4) temporary evacuation; (5) long term or permanent emigration. At the specification of zones of planned protection it is also coming out that regulation of movement of persons, sheltering and iodine prophylaxis were ordered in advance based on the evaluation of the crashed establishment state. In such situation the decision on protective measures in the time interval 6 to 12 hours after the beginning of accidental release is forwarding to: withdrawing the accepted orders on measures and transition from sheltering to temporary evacuation. The criterion for temporary evacuation is: (1) probability of exceeding the effective dose 100 mSv for children up to 10 years of age and pregnant women and 500 mSv for other population within 48 hours after beginning of accidental release; (2) probability of averting the effective dose 50 mSv up to 7 days, 100 mSv up to 15 days and 150 mSv up to 30 days for all population groups. In next part the intervention level, interpretation of values of kerma dose rate in air and determination of the size of planned protection zones are discussed. (J.K.) 3 tabs.

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

    International Nuclear Information System (INIS)

    Genter, G.

    1981-03-01

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

  18. Impact of sophisticated fog spray models on accident analyses

    International Nuclear Information System (INIS)

    Roblyer, S.P.; Owzarski, P.C.

    1978-01-01

    The N-Reactor confinement system release dose to the public in a postulated accident is reduced by washing the confinement atmosphere with fog sprays. This allows a low pressure release of confinement atmosphere containing fission products through filters and out an elevated stack. The current accident analysis required revision of the CORRAL code and other codes such as CONTEMPT to properly model the N Reactor confinement into a system of multiple fog-sprayed compartments. In revising these codes, more sophisticated models for the fog sprays and iodine plateout were incorporated to remove some of the conservatism of steam condensing rate, fission product washout and iodine plateout than used in previous studies. The CORRAL code, which was used to describe the transport and deposition of airborne fission products in LWR containment systems for the Rasmussen Study, was revised to describe fog spray removal of molecular iodine (I 2 ) and particulates in multiple compartments for sprays having individual characteristics of on-off times, flow rates, fall heights, and drop sizes in changing containment atmospheres. During postulated accidents, the code determined the fission product removal rates internally rather than from input decontamination factors. A discussion is given of how the calculated plateout and washout rates vary with time throughout the analysis. The results of the accident analyses indicated that more credit could be given to fission product washout and plateout. An important finding was that the release of fission products to the atmosphere and adsorption of fission products on the filters were significantly lower than previous studies had indicated

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

    Energy Technology Data Exchange (ETDEWEB)

    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

  20. ESTER: a new approach in modelling severe accidents

    International Nuclear Information System (INIS)

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

    1993-01-01

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

  1. Synthesis of the models used in France for the evaluation of the consequences of accident

    International Nuclear Information System (INIS)

    Crabol, B.

    1992-01-01

    In order to evaluate the consequences of an atmospheric release in case of an accident on a nuclear installation, different predictive models have been developed by the organizations involved in the management of the crisis. These models are of different numerical complexity: precalculated graphs, gaussian puff models or 3D models. The harmonization of these models, the definition of their use, notably in the first phases of the accident (predictive and real-time phases) have been discussed in a working group including representants of the utility, the safety authorities and the Meteorological Office. The reflexions of the group, the models already operational, those still under discussion and their use in the different technical crisis centers are presented

  2. Input-output model for MACCS nuclear accident impacts estimation¹

    Energy Technology Data Exchange (ETDEWEB)

    Outkin, Alexander V. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Bixler, Nathan E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Vargas, Vanessa N [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2015-01-27

    Since the original economic model for MACCS was developed, better quality economic data (as well as the tools to gather and process it) and better computational capabilities have become available. The update of the economic impacts component of the MACCS legacy model will provide improved estimates of business disruptions through the use of Input-Output based economic impact estimation. This paper presents an updated MACCS model, bases on Input-Output methodology, in which economic impacts are calculated using the Regional Economic Accounting analysis tool (REAcct) created at Sandia National Laboratories. This new GDP-based model allows quick and consistent estimation of gross domestic product (GDP) losses due to nuclear power plant accidents. This paper outlines the steps taken to combine the REAcct Input-Output-based model with the MACCS code, describes the GDP loss calculation, and discusses the parameters and modeling assumptions necessary for the estimation of long-term effects of nuclear power plant accidents.

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

    International Nuclear Information System (INIS)

    Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

    1997-01-01

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

  4. The development of a nuclear accident risk information system

    International Nuclear Information System (INIS)

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

    2001-01-01

    The computerized system NARIS (Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analyses of the distribution of the health effects. The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system

  5. [Guilty victims: a model to perpetuate impunity for work-related accidents].

    Science.gov (United States)

    Vilela, Rodolfo Andrade Gouveia; Iguti, Aparecida Mari; Almeida, Ildeberto Muniz

    2004-01-01

    This article analyzes reports and data from the investigation of severe and fatal work-related accidents by the Regional Institute of Criminology in Piracicaba, São Paulo State, Brazil. Some 71 accident investigation reports were analyzed from 1998, 1999, and 2000. Accidents involving machinery represented 38.0% of the total, followed by high falls (15.5%), and electric shocks (11.3%). The reports conclude that 80.0% of the accidents are caused by "unsafe acts" committed by workers themselves, while the lack of safety or "unsafe conditions" account for only 15.5% of cases. Victims are blamed even in situations involving high risk in which not even minimum safety conditions are adopted, thus favoring employers' interests. Such conclusions reflect traditional reductionist explanatory models, in which accidents are viewed as simple, unicausal phenomena, generally focused on slipups and errors by the workers themselves. Despite criticism in recent decades from the technical and academic community, this concept is still hegemonic, thus jeopardizing the development of preventive policies and the improvement of work conditions.

  6. International collaboration for development of accident-resistant LWR fuel. International Collaboration for Development of Accident Resistant Light Water Reactor Fuel

    International Nuclear Information System (INIS)

    Sowder, Andrew

    2013-01-01

    Department of Energy is providing substantial support for initial R and D on accident-tolerant fuel concepts with an aggressive target of a lead test assembly (LTA) in an LWR by 2022. EPRI proposes an additional stretch goal of commercialisation of a new LWR fuel by 2030. The scale of and resource demands associated with these R and D targets require a global collaborative structure to leverage resources, create an environment for innovation and co-operation, and foster necessary partnerships and arrangements among the many key players and roles spanning government, academic, and industrial sectors. EPRI is proposing a voluntary, open, and non-binding structure to quickly build momentum and to maximise early engagement and information exchange among key stakeholders. The flexibility of this organisational model offers an environment that is compatible with and encourages engagement, innovation, and development of the more formal arrangements and partnerships that will be needed to commercialise current R and D concepts. The opportunity for transformation of LWR fuel performance under normal and accident conditions is now. Accordingly, the time for action is now. Commercialisation of accident-tolerant fuel in the near future can only be realised with collaboration among governments, industry and academia on a scale commensurate with the challenges at hand

  7. Health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Evans, J.S.; Abrahmson, S.; Bender, M.A.; Boecker, B.B.; Scott, B.R.; Gilbert, E.S.

    1993-10-01

    This report is a revision of NUREG/CR-4214, Rev. 1, Part 1 (1990), Health Effects Models for Nuclear Power Plant Accident Consequence Analysis. This revision has been made to incorporate changes to the Health Effects Models recommended in two addenda to the NUREG/CR-4214, Rev. 1, Part 11, 1989 report. The first of these addenda provided recommended changes to the health effects models for low-LET radiations based on recent reports from UNSCEAR, ICRP and NAS/NRC (BEIR V). The second addendum presented changes needed to incorporate alpha-emitting radionuclides into the accident exposure source term. As in the earlier version of this report, models are provided for early and continuing effects, cancers and thyroid nodules, and genetic effects. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes are considered. Linear and linear-quadratic models are recommended for estimating the risks of seven types of cancer in adults - leukemia, bone, lung, breast, gastrointestinal, thyroid, and ''other''. For most cancers, both incidence and mortality are addressed. Five classes of genetic diseases -- dominant, x-linked, aneuploidy, unbalanced translocations, and multifactorial diseases are also considered. Data are provided that should enable analysts to consider the timing and severity of each type of health risk

  8. Testing of an accident consequence assessment model using field data

    International Nuclear Information System (INIS)

    Homma, Toshimitsu; Matsubara, Takeshi; Tomita, Kenichi

    2007-01-01

    This paper presents the results obtained from the application of an accident consequence assessment model, OSCAAR to the Iput dose reconstruction scenario of BIOMASS and also to the Chernobyl 131 I fallout scenario of EMRAS, both organized by International Atomic Energy Agency. The Iput Scenario deals with 137 Cs contamination of the catchment basin and agricultural area in the Bryansk Region of Russia, which was heavily contaminated after the Chernobyl accident. This exercise was used to test the chronic exposure pathway models in OSCAAR with actual measurements and to identify the most important sources of uncertainty with respect to each part of the assessment. The OSCAAR chronic exposure pathway models had some limitations but the refined model, COLINA almost successfully reconstructed the whole 10-year time course of 137 Cs activity concentrations in most requested types of agricultural products and natural foodstuffs. The Plavsk scenario provides a good opportunity to test not only the food chain transfer model of 131 I but also the method of assessing 131 I thyroid burden. OSCAAR showed in general good capabilities for assessing the important 131 I exposure pathways. (author)

  9. Modelling and assessment of accident consequences: Development of a computer-assisted decision-support system RODOS/RESY for nuclear emergencies

    International Nuclear Information System (INIS)

    Benz, G.; Ehrhardt, J.; Paesler-Sauer, J.; Schichtel, T.; Schuele, O.; Steinhauer, C.

    1995-01-01

    In cooperation with NRPB, the specifications of the mainframe COSYMA version 95/1 and the PC COSYMA version 2.0 were prepared and the corresponding modifications implemented. Important improvements are dose-rate dependent models for deterministic health effects, the time dependent efficiency of stable iodine tablets, the extension of data bases for the inclusion of activation products, and supplementary evaluation programs. PC COSYMA has been completed by an economics module, further options in the ingestion pathways, and a graphics package for presenting assessment results. COSYMA has been applied for probabilistic dose assessments within paramter studies and special investigations of EPR concepts. RODOS, the real-time on-line decision support system for nuclear emergency management, has been further developed with the aim of the first pilot version 2.0 for pre-operational application in the second half of 1995. At present, some 20 institutes in the EU, 8 institutes in Russia, Belarus and Ukraine, and 5 institutes in central and eastern European countries cooperate with FZKA/INR as main coordinator. The working meeting of all contractors in Interlaken (CH), 20-24 June 1994, lead to the consolidation of the international cooperation and a broad consensus about the future role and use of the RODOS-system, in particular in the Eastern countries. At the national level, the approval of a strategy and position paper on RODOS/RESY (the near range and early phase subsystem) by subcommittees of the German Radiation Protection Commission can be considered as a step further to the future use of the system in emergency centres of the Federal States. (orig./HP)

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

    Directory of Open Access Journals (Sweden)

    Robert P. Martin

    2012-01-01

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

  11. Development of supporting system for emergency response to maritime transport accidents involving radioactive material

    International Nuclear Information System (INIS)

    Odano, N.; Matsuoka, T.; Suzuki, H.

    2004-01-01

    National Maritime Research Institute has developed a supporting system for emergency response of competent authority to maritime transport accidents involving radioactive material. The supporting system for emergency response has functions of radiation shielding calculation, marine diffusion simulation, air diffusion simulation and radiological impact evaluation to grasp potential hazard of radiation. Loss of shielding performance accident and loss of sealing ability accident were postulated and impact of the accidents was evaluated based on the postulated accident scenario. Procedures for responding to emergency were examined by the present simulation results

  12. Operator modeling of a loss-of-pumping accident using MicroSAINT

    International Nuclear Information System (INIS)

    Olsen, L.M.

    1992-01-01

    The Savannah River Laboratory (SRL) human factors group has been developing methods for analyzing nuclear reactor operator actions during hypothetical design-basis accident scenarios. The SRL reactors operate at a lower temperature and pressure than power reactors resulting in accident sequences that differ from those of power reactors. Current methodology development is focused on modeling control room operator response times dictated by system event times specified in the Savannah River Site Reactor Safety Analysis Report (SAR). The modeling methods must be flexible enough to incorporate changes to hardware, procedures, or postulated system event times and permit timely evaluation. The initial model developed was for the loss-of-pumping accident (LOPA) because a significant number of operator actions are required to respond to this postulated event. Human factors engineers had been researching and testing a network modeling simulation language called MicroSAINT to simulate operators' personal and interpersonal actions relative to operating system events. The LOPA operator modeling project demonstrated the versatility and flexibility of MicroSAINT for modeling control room crew interactions

  13. Accident management

    International Nuclear Information System (INIS)

    Lutz, R.J.; Monty, B.S.; Liparulo, N.J.; Desaedeleer, G.

    1989-01-01

    The foundation of the framework for a Severe Accident Management Program is the contained in the Probabilistic Safety Study (PSS) or the Individual Plant Evaluations (IPE) for a specific plant. The development of a Severe Accident Management Program at a plant is based on the use of the information, in conjunction with other applicable information. A Severe Accident Management Program must address both accident prevention and accident mitigation. The overall Severe Accident Management framework must address these two facets, as a living program in terms of gathering the evaluating information, the readiness to respond to an event. Significant international experience in the development of severe accident management programs exist which should provide some direction for the development of Severe Accident Management in the U.S. This paper reports that the two most important elements of a Severe Accident Management Program are the Emergency Consultation process and the standards for measuring the effectiveness of individual Severe Accident Management Programs at utilities

  14. Critical analysis of accident scenario and consequences modelling applied to light-water reactor power plants for accident categories beyond the design basis accident (DBA)

    International Nuclear Information System (INIS)

    Brofferio, C.; Cagnetti, P.; Ferrara, V.; Manilia, E.; Pietrangeli, G.; Sennis, C.

    1985-01-01

    A critical analysis and sensitivity study of the modelling of accident scenarios and environmental consequences are presented, for light-water reactor accident categories beyond the standard design-basis-accident category. The first chapter, on ''source term'' deals with the release of fission products from a damaged core inventory and their migration within the primary circuit and the reactor containment. Particular attention is given to the influence of engineering safeguards intervention and of the chemical forms of the released fission products. The second chapter deals with their release to the atmosphere, transport and wet or dry deposition, outlining relevant partial effects and confronting short-duration or prolonged releases. The third chapter presents a variability analysis, for environmental contamination levels, for two extreme hypothetical scenarios, evidencing the importance of plume rise. A numerical plume rise model is outlined

  15. Comparison of US/FRG accident condition models for HTGR fuel failure and radionuclide release

    International Nuclear Information System (INIS)

    Verfondern, K.

    1991-03-01

    The objective was to compare calculation models used in safety analyses in the US and FRG which describe fission product release behavior from TRISO coated fuel particles under core heatup accident conditions. The frist step performed is the qualitative comparison of both sides' fuel failure and release models in order to identify differences and similarities in modeling assumptions and inputs. Assumptions of possible particle failure mechanisms under accident conditions (SiC degradation, pressure vessel) are principally the same on both sides though they are used in different modeling approaches. The characterization of a standard (= intact) coated particle to be of non-releasing (GA) or possibly releasing (KFA/ISF) type is one of the major qualitative differences. Similar models are used regarding radionuclide release from exposed particle kernels. In a second step, a quantitative comparison of the calculation models was made by assessing a benchmark problem predicting particle failure and radionuclide release under MHTGR conduction cooldown accident conditions. Calculations with each side's reference method have come to almost the same failure fractions after 250 hours for the core region with maximum core heatup temperature despite the different modeling approaches of SORS and PANAMA-I. The comparison of the results of particle failure obtained with the Integrated Failure and Release Model for Standard Particles and its revision provides a 'verification' of these models in this sense that the codes (SORS and PANAMA-II, and -III, respectively) which were independently developed lead to very good agreement in the predictions. (orig./HP) [de

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

    International Nuclear Information System (INIS)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Young Ro and others

    2000-07-01

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

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

    Directory of Open Access Journals (Sweden)

    Rackynelly Alves Sarmento Soares

    2013-01-01

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

  18. Modelling transport and deposition of caesium and iodine from the Chernobyl accident using the DREAM model

    Directory of Open Access Journals (Sweden)

    J. Brandt

    2002-01-01

    Full Text Available A tracer model, DREAM (the Danish Rimpuff and Eulerian Accidental release Model, has been developed for modelling transport, dispersion and deposition (wet and dry of radioactive material from accidental releases, as the Chernobyl accident. The model is a combination of a Lagrangian model, that includes the near source dispersion, and an Eulerian model describing the long-range transport. The performance of the transport model has previously been tested within the European Tracer Experiment, ETEX, which included transport and dispersion of an inert, non-depositing tracer from a controlled release. The focus of this paper is the model performance with respect to the total deposition of  137Cs, 134Cs and 131I from the Chernobyl accident, using different relatively simple and comprehensive parameterizations for dry- and wet deposition. The performance, compared to measurements, of using different combinations of two different wet deposition parameterizations and three different parameterizations of dry deposition has been evaluated, using different statistical tests. The best model performance, compared to measurements, is obtained when parameterizing the total deposition combined of a simple method for dry deposition and a subgrid-scale averaging scheme for wet deposition based on relative humidities. The same major conclusion is obtained for all the three different radioactive isotopes and using two different deposition measurement databases. Large differences are seen in the results obtained by using the two different parameterizations of wet deposition based on precipitation rates and relative humidities, respectively. The parameterization based on subgrid-scale averaging is, in all cases, performing better than the parameterization based on precipitation rates. This indicates that the in-cloud scavenging process is more important than the below cloud scavenging process for the submicron particles and that the precipitation rates are

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

    International Nuclear Information System (INIS)

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

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided

  20. The accident consequence model of the German safety study

    International Nuclear Information System (INIS)

    Huebschmann, W.

    1977-01-01

    The accident consequence model essentially describes a) the diffusion in the atmosphere and deposition on the soil of radioactive material released from the reactor into the atmosphere; b) the irradiation exposure and health consequences of persons affected. It is used to calculate c) the number of persons suffering from acute or late damage, taking into account possible counteractions such as relocation or evacuation, and d) the total risk to the population from the various types of accident. The model, the underlying parameters and assumptions are described. The bone marrow dose distribution is shown for the case of late overpressure containment failure, which is discussed in the paper of Heuser/Kotthoff, combined with four typical weather conditions. The probability distribution functions for acute mortality, late incidence of cancer and genetic damage are evaluated, assuming a characteristic population distribution. The aim of these calculations is first the presentation of some results of the consequence model as an example, in second the identification of problems, which need possibly in a second phase of study to be evaluated in more detail. (orig.) [de

  1. Development of LWR Fuels with Enhanced Accident Tolerance

    International Nuclear Information System (INIS)

    Lahoda, Edward J.; Boylan, Frank A.

    2015-01-01

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company's Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U 15 N and U 3 Si 2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U 3 Si 2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U 3 Si 2 will represent a 15% increase in U235 loadings over those in UO fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U 3 Si 2 /68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO 2 pellets. Pellets and powders of UO 2 , UN, and U 3 Si 2 the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO 2 . Cold spray application of either the amorphous steel or the Ti 2 AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the

  2. Development of LWR Fuels with Enhanced Accident Tolerance

    Energy Technology Data Exchange (ETDEWEB)

    Lahoda, Edward J. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States); Boylan, Frank A. [Westinghouse Electric Company, LLC, Cranberry Woods, PA (United States)

    2015-10-30

    Significant progress was made on the technical, licensing, and business aspects of the Westinghouse Electric Company’s Enhanced Accident Tolerant Fuel (ATF) by the Westinghouse ATF team. The fuel pellet options included waterproofed U15N and U3Si2 and the cladding options SiC composites and zirconium alloys with surface treatments. Technology was developed that resulted in U3Si2 pellets with densities of >94% being achieved at the Idaho National Laboratory (INL). The use of U3Si2 will represent a 15% increase in U235 loadings over those in UO₂ fuel pellets. This technology was then applied to manufacture pellets for 6 test rodlets which were inserted in the Advanced Test Reactor (ATR) in early 2015 in zirconium alloy cladding. The first of these rodlets are expected to be removed in about 2017. Key characteristics to be determined include verification of the centerline temperature calculations, thermal conductivity, fission gas release, swelling and degree of amorphization. Waterproofed UN pellets have achieved >94% density for a 32% U3Si2/68% UN composite pellet at Texas A&M University. This represents a U235 increase of about 31% over current UO2 pellets. Pellets and powders of UO2, UN, and U3Si2the were tested by Westinghouse and Los Alamos National Laboratory (LANL) using differential scanning calorimetry to determine what their steam and 20% oxygen corrosion temperatures were as compared to UO2. Cold spray application of either the amorphous steel or the Ti2AlC was successful in forming an adherent ~20 micron coating that remained after testing at 420°C in a steam autoclave. Tests at 1200°C in 100% steam on coatings for Zr alloy have not been successful, possibly due to the low density of the coatings which allowed steam transport to the base zirconium metal. Significant modeling and testing

  3. Global atmospheric dispersion modelling after the Fukushima accident

    Energy Technology Data Exchange (ETDEWEB)

    Suh, K.S.; Youm, M.K.; Lee, B.G.; Min, B.I. [Korea Atomic Energy Research Institute (Korea, Republic of); Raul, P. [Universidad de Sevilla (Spain)

    2014-07-01

    A large amount of radioactive material was released to the atmosphere due to the Fukushima nuclear accident in March 2011. The radioactive materials released into the atmosphere were mostly transported to the Pacific Ocean, but some of them were fallen on the surface due to dry and wet depositions in the northwest area from the Fukushima nuclear site. Therefore, northwest part of the nuclear site was seriously contaminated and it was designated with the restricted zone within a radius of 20 ∼ 30 km around the Fukushima nuclear site. In the early phase of the accident from 11 March to 30 March, the radioactive materials were dispersed to an area of the inland and offshore of the nuclear site by the variations of the wind. After the Fukushima accident, the radionuclides were detected through the air monitoring in the many places over the world. The radioactive plume was transported to the east part off the site by the westerly jet stream. It had detected in the North America during March 17-21, in European countries during March 23-24, and in Asia during from March 24 to April 6, 2011. The radioactive materials were overall detected across the northern hemisphere passed by 15 ∼ 20 days after the accident. Three dimensional numerical model was applied to evaluate the dispersion characteristics of the radionuclides released into the air. Simulated results were compared with measurements in many places over the world. Comparative results had good agreements in some places, but they had a little differences in some locations. The difference between the calculations and measurements are due to the meteorological data and relatively coarse resolutions in the model. Some radioactive materials were measured in Philippines, Taiwan, Hon Kong and South Korea during from March 23-28. It inferred that it was directly transported from the Fukushima by the northeastern monsoon winds. This event was well represented in the numerical model. Generally, the simulations had a good

  4. Utilization of dose assessment models to facilitate off-site recovery operations for accidents at nuclear facilities

    International Nuclear Information System (INIS)

    Dickerson, M.H.; Foster, K.T.

    1989-09-01

    One of the most important uses of dose assessment models in response to accidents at nuclear facilities is to help provide guidance to emergency response managers for identifying, and mitigating, the consequences of an accident once the accident has been terminated. By combining results from assessment models with radiological measurements, a qualitative methodology can be developed to aid emergency response managers in determining the total dose received by the population and to minimize future doses through the use of mitigation procedures. To illustrate the methodology, this discussion focuses on the use of models to estimate the dose delivered to the public both during and after a nuclear accident. 4 refs., 10 figs., 1 tab

  5. Unavoidable Accident

    OpenAIRE

    Grady, Mark F.

    2009-01-01

    In negligence law, "unavoidable accident" is the risk that remains when an actor has used due care. The counterpart of unavoidable accident is "negligent harm." Negligence law makes parties immune for unavoidable accident even when they have used less than due care. Courts have developed a number of methods by which they "sort" accidents to unavoidable accident or to negligent harm, holding parties liable only for the latter. These sorting techniques are interesting in their own right and als...

  6. Development of the methodology and approaches to validate safety and accident management

    International Nuclear Information System (INIS)

    Asmolov, V.G.

    1997-01-01

    The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987-1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed. (orig.)

  7. AN EFFECTIVE RISK-PREVENTIVE MODEL PROPOSAL FOR OCCUPATIONAL ACCIDENTS AT SHIPYARDS

    Directory of Open Access Journals (Sweden)

    Ozge Acuner

    2016-03-01

    Full Text Available According to the statistics of occupational accidents, it is observed that the number of accidents occurred in shipbuilding industry is high and the rate of deaths and serious injuries among these accidents is higher than in other industries. However, the number of the studies to prevent these accidents in both industrial and scientific practices is considerably low. Therefore, the objective of this study is to develop an efficient risk preventive model in accordance with occupational health and safety regulations for industrial organizations. The approach proposed in this study differs from those described in the literature, because it is based on fuzzy set theory in order to cope with uncertainties on probability and severity definitions in terms of occupational health and safety. Furthermore, in this paper, risk severity is considered in terms of harm to worker, harm to environment, and harm to hardware, whereas in the literature, risk severity is generally considered solely in terms of only harm to worker. Then, risk magnitude is obtained by utilizing fuzzy inference system. The proposed approach is applied to a shipyard located in the Marmara Region in order to illustrate the applicability of the model.

  8. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    International Nuclear Information System (INIS)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha

    2016-01-01

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  9. Code Development on Fission Product Behavior under Severe Accident-Validation of Aerosol Sedimentation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Jang, Jin Sung; Kim, Dong Ha [KAERI, Daejeon (Korea, Republic of)

    2016-05-15

    The gas and aerosol phases of the radioactive materials move through the reactor coolant systems and containments as loaded on the carrier gas or liquid, such as steam or water. Most radioactive materials might escape in the form of aerosols from a nuclear power plant during a severe reactor accident, and it is very important to predict the behavior of these radioactive aerosols in the reactor cooling system and in the containment building under severe accident conditions. Aerosols are designated as very small solid particles or liquid droplets suspended in a gas phase. The suspended solid or liquid particles typically have a range of sizes of 0.01 m to 20 m. Aerosol concentrations in reactor accident analyses are typically less than 100 g/m3 and usually less than 1 g/m3. When there are continuing sources of aerosol to the gas phase or when there are complicated processes involving engineered safety features, much more complicated size distributions develop. It is not uncommon for aerosols in reactor containments to have bimodal size distributions for at least some significant periods of time early during an accident. Salient features of aerosol physics under reactor accident conditions that will affect the nature of the aerosols are (1) the formation of aerosol particles, (2) growth of aerosol particles, (3) shape of aerosol particles. At KAERI, a fission product module has been developed to predict the behaviors of the radioactive materials in the reactor coolant system under severe accident conditions. The fission product module consists of an estimation of the initial inventories, species release from the core, aerosol generation, gas transport, and aerosol transport. The final outcomes of the fission product module designate the radioactive gas and aerosol distribution in the reactor coolant system. The aerosol sedimentation models in the fission product module were validated using ABCOVE and LACE experiments. There were some discrepancies on the predicted

  10. Coolability in the frame of core melt accidents in light water reactors. Model development and validation for ATHLET-CD and ASTEC. Final report; Kuehlbarkeit im Rahmen von Kernschmelzunfaellen bei Leichtwasserreaktoren. Modellentwicklung und Validierung fuer ATHLET-CD und ASTEC. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buck, Michael; Pohlner, Georg; Rahman, Saidur; Berkhan, Ana

    2015-07-15

    The code system ATHLET/ATHLET-CD is being developed in the frame of the reactor safety research of the German Federal Ministry for Economic Affairs and Energy (BMWi) within the topic analysis of transients and accident sequences. It serves for simulation of transients and accidents to be used in safety analyses for light water reactors. In the present project the development and validation of models for ATHLET-CD for description of the processes during severe accidents are continued. These works should enable broad safety analyses by a mechanistic description of the processes even during late phases of a degrading core and by this a profound estimation on coolability and accident management options during every phase. With the actual status of modelling in ATHLET-CD analyses on coolability are made to give a solid base for estimates about stabilization by cooling or accident progression, dependent on the scenario. The modeling in the MEWA module, describing the processes in a severely degraded core in ATHLET-CD, is extended on the processes in the lower plenum. For this, the model on melt pool behavior is extended and linked to the RPV wall. The coupling between MEWA and the thermal-hydraulics of ATHLET-CD is improved. The validation of the models is continued by calculations on new experiments and comparing analyses done in the frame of the European Network SARNET-2. For the European integral code ASTEC contributions from the modeling for ATHLET-CD will be done, especially by providing a model for the melt behavior in the lower plenum of a LWR. This report illustrates the work carried out in the frame of this project, and shows results of calculations and the status of validation by recalculations on experiments for debris bed coolability, melt pool behavior as well as jet fragmentation and debris bed formation.

  11. ASTEC code development, validation and applications for severe accident management within the CESAM European project - 15392

    International Nuclear Information System (INIS)

    Van Dorsselaere, J.P.; Chatelard, P.; Chevalier-Jabet, K.; Nowack, H.; Herranz, L.E.; Pascal, G.; Sanchez-Espinoza, V.H.

    2015-01-01

    ASTEC, jointly developed by IRSN and GRS, is considered as the European reference code since it capitalizes knowledge from the European research on the domain. The CESAM project aims at its enhancement and extension for use in severe accident management (SAM) analysis of the nuclear power plants (NPP) of Generation II-III presently under operation or foreseen in near future in Europe, spent fuel pools included. Within the CESAM project 3 main types of research activities are performed: -) further validation of ASTEC models important for SAM, in particular for the phenomena being of importance in the Fukushima-Daichi accidents, such as reflooding of degraded cores, pool scrubbing, hydrogen combustion, or spent fuel pools behaviour; -) modelling improvements, especially for BWR or based on the feedback of validation tasks; and -) ASTEC applications to severe accident scenarios in European NPPs in order to assess prevention and mitigation measures. An important step will be reached with the next major ASTEC V2.1 version planned to be delivered in the first part of 2015. Its main improvements will concern the possibility to simulate in details the core degradation of BWR and PHWR and a model of reflooding of severely degraded cores. A new user-friendly Graphic User Interface will be available for plant analyses

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

    International Nuclear Information System (INIS)

    Chatelard, P.; Reinke, N.; Arndt, S.; Belon, S.; Cantrel, L.; Carenini, L.; Chevalier-Jabet, K.; Cousin, F.; Eckel, J.; Jacq, F.; Marchetto, C.; Mun, C.; Piar, L.

    2014-01-01

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

  13. Development of methodology for the analysis of fuel behavior in light water reactor in design basis accidents

    International Nuclear Information System (INIS)

    Salatov, A. A.; Goncharov, A. A.; Eremenko, A. S.; Kuznetsov, V. I.; Bolnov, V. A.; Gusev, A. S.; Dolgov, A. B.; Ugryumov, A. V.

    2013-01-01

    The report attempts to analyze the current experience of the safety fuel for light-water reactors (LWRs) under design-basis accident conditions in terms of its compliance with international requirements for licensing nuclear power plants. The components of fuel behavior analysis methodology in design basis accidents in LWRs were considered, such as classification of design basis accidents, phenomenology of fuel behavior in design basis accidents, system of fuel safety criteria and their experimental support, applicability of used computer codes and input data for computational analysis of the fuel behavior in accidents, way of accounting for the uncertainty of calculation models and the input data. A brief history of the development of probabilistic safety analysis methodology for nuclear power plants abroad is considered. The examples of a conservative approach to safety analysis of VVER fuel and probabilistic approach to safety analysis of fuel TVS-K are performed. Actual problems in development of the methodology of analyzing the behavior of VVER fuel at the design basis accident conditions consist, according to the authors opinion, in following: 1) Development of a common methodology for analyzing the behavior of VVER fuel in the design basis accidents, implementing a realistic approach to the analysis of uncertainty - in the future it is necessary for the licensing of operating VVER fuel abroad; 2) Experimental and analytical support to the methodology: experimental studies to identify and study the characteristics of the key uncertainties of computational models of fuel and the cladding, development of computational models of key events in codes, validation code on the basis of integral experiments

  14. Complex accident scenarios modelled and analysed by Stochastic Petri Nets

    International Nuclear Information System (INIS)

    Nývlt, Ondřej; Haugen, Stein; Ferkl, Lukáš

    2015-01-01

    This paper is focused on the usage of Petri nets for an effective modelling and simulation of complicated accident scenarios, where an order of events can vary and some events may occur anywhere in an event chain. These cases are hardly manageable by traditional methods as event trees – e.g. one pivotal event must be often inserted several times into one branch of the tree. Our approach is based on Stochastic Petri Nets with Predicates and Assertions and on an idea, which comes from the area of Programmable Logic Controllers: an accidental scenario is described as a net of interconnected blocks, which represent parts of the scenario. So the scenario is firstly divided into parts, which are then modelled by Petri nets. Every block can be easily interconnected with other blocks by input/output variables to create complex ones. In the presented approach, every event or a part of a scenario is modelled only once, independently on a number of its occurrences in the scenario. The final model is much more transparent then the corresponding event tree. The method is shown in two case studies, where the advanced one contains a dynamic behavior. - Highlights: • Event & Fault trees have problems with scenarios where an order of events can vary. • Paper presents a method for modelling and analysis of dynamic accident scenarios. • The presented method is based on Petri nets. • The proposed method solves mentioned problems of traditional approaches. • The method is shown in two case studies: simple and advanced (with dynamic behavior)

  15. Modeling and analyses of postulated UF6 release accidents in gaseous diffusion plant

    International Nuclear Information System (INIS)

    Kim, S.H.; Taleyarkhan, R.P.; Keith, K.D.; Schmidt, R.W.; Carter, J.C.; Dyer, R.H.

    1995-10-01

    Computer models have been developed to simulate the transient behavior of aerosols and vapors as a result of a postulated accident involving the release of uranium hexafluoride (UF 6 ) into the process building of a gaseous diffusion plant. UF 6 undergoes an exothermic chemical reaction with moisture (H 2 O) in the air to form hydrogen fluoride (HF) and radioactive uranyl fluoride (UO 2 F 2 ). As part of a facility-wide safety evaluation, this study evaluated source terms consisting of UO 2 F 2 as well as HF during a postulated UF 6 release accident in a process building. In the postulated accident scenario, ∼7900 kg (17,500 lb) of hot UF 6 vapor is released over a 5 min period from the process piping into the atmosphere of a large process building. UO 2 F 2 mainly remains as airborne-solid particles (aerosols), and HF is in a vapor form. Some UO 2 F 2 aerosols are removed from the air flow due to gravitational settling. The HF and the remaining UO 2 F 2 are mixed with air and exhausted through the building ventilation system. The MELCOR computer code was selected for simulating aerosols and vapor transport in the process building. MELCOR model was first used to develop a single volume representation of a process building and its results were compared with those from past lumped parameter models specifically developed for studying UF 6 release accidents. Preliminary results indicate that MELCOR predicted results (using a lumped formulation) are comparable with those from previously developed models

  16. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    International Nuclear Information System (INIS)

    Wang, Jun; Mccabe, Mckinleigh; Wu, Lei; Dong, Xiaomeng; Wang, Xianmao; Haskin, Troy Christopher; Corradini, Michael L.

    2017-01-01

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  17. Accident tolerant clad material modeling by MELCOR: Benchmark for SURRY Short Term Station Black Out

    Energy Technology Data Exchange (ETDEWEB)

    Wang, Jun, E-mail: jwang564@wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Mccabe, Mckinleigh [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wu, Lei [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Dong, Xiaomeng [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Wang, Xianmao [Institute of Nuclear and New Energy Technology, Tsinghua University, Beijing 100084 (China); Haskin, Troy Christopher [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States); Corradini, Michael L., E-mail: corradini@engr.wisc.edu [College of Engineering, The University of Wisconsin-Madison, Madison 53706 (United States)

    2017-03-15

    Highlights: • Thermo-physical and oxidation kinetics properties calculation and analysis of FeCrAl. • Properties modelling of FeCrAl in MELCOR. • Benchmark calculation of Surry nuclear power plant. - Abstract: Accident tolerant fuel and cladding materials are being investigated to provide a greater resistance to fuel degradation, oxidation and melting if long-term cooling is lost in a Light Water Reactor (LWR) following an accident such as a Station Blackout (SBO) or Loss of Coolant Accident (LOCA). Researchers at UW-Madison are analyzing an SBO sequence and examining the effect of a loss of auxiliary feedwater (AFW) with the MELCOR systems code. Our research work considers accident tolerant cladding materials (e.g., FeCrAl alloy) and their effect on the accident behavior. We first gathered the physical properties of this alternative cladding material via literature review and compared it to the usual zirconium alloys used in LWRs. We then developed a model for the Surry reactor for a Short-term SBO sequence and examined the effect of replacing FeCrAl for Zircaloy cladding. The analysis uses MELCOR, Version 1.8.6 YR, which is developed by Idaho National Laboratory in collaboration with MELCOR developers at Sandia National Laboratories. This version allows the user to alter the cladding material considered, and our study examines the behavior of the FeCrAl alloy as a substitute for Zircaloy. Our benchmark comparisons with the Sandia National Laboratory’s analysis of Surry using MELCOR 1.8.6 and the more recent MELCOR 2.1 indicate good overall agreement through the early phases of the accident progression. When FeCrAl is substituted for Zircaloy to examine its performance, we confirmed that FeCrAl slows the accident progression and reduce the amount of hydrogen generated. Our analyses also show that this special version of MELCOR can be used to evaluate other potential ATF cladding materials, e.g., SiC as well as innovative coatings on zirconium cladding

  18. A note on modeling road accident frequency: a flexible elasticity model.

    Science.gov (United States)

    Couto, António; Ferreira, Sara

    2011-11-01

    Count data models and their variants have been widely applied in accident modeling. The traditional log-linear function is used to represent the relationship between explanatory variables and the dependent variable (accident frequency). However, this function assumes constant elasticity for the estimation parameters, which is a limitation in the analysis of the effects of explanatory variables on accident risk. Although interaction effects between explanatory variables have been studied in the road safety context (where they are normally assessed by logistic regression), no one has yet examined the possibility of using a flexible function form allowing non-constant elasticity values. This paper seeks to explore the use of the translog function usually used in the economics context to allow the elasticity to vary with the values of other explanatory variables. Therefore, the objective of this study was to evaluate the application of the translog function to accident modeling and to compare the results with those of the traditional log-linear function negative binomial (NB) model. The results show that, in terms of goodness-of-fit statistics and residual analysis, the NB model with the translog function performs better than the traditional NB model. Additional evaluations in terms of predictive performance, hotspot identification and uncertainty associated with the estimated values were taken into account. Although this study is exploratory in nature, it suggests that the translog function has considerable potential for modeling accident observations. It is hoped that this novel accident modeling methodology will open the door to the reliable interpretation and evaluation of the influence of explanatory variables on accident frequency. Copyright © 2011 Elsevier Ltd. All rights reserved.

  19. A Study of The Relationship Between The Components of The Five-Factor Model of Personality and The Occurrence of Occupational Accidents in Industry Workers

    Directory of Open Access Journals (Sweden)

    Ehsanollah Habibi

    2016-05-01

    Full Text Available Accidents are among the most important problems of both the developed and the developing countries. Individual factors and personality traits are the primary causes of human errors and contribute to accidents. The present study aims to investigate the relationship between the components of the five-factor model of personality and the occurrence of occupational accidents in industrial workers. The independent T-test indicated that there is a meaningful relationship between the personality traits and accident proneness. In the two groups of industry workers injured in occupational accidents and industry workers without any occupational accidents, there is a significant relationship between personality traits, neuroticism (p=0.001, openness to experience (p=0.001, extraversion (p=0.024 and conscientiousness (p=0.021. Nonetheless, concerning the personality trait of agreeableness (p = 0.09, the group of workers with accidents did not differ significantly from the workers without any accidents. The results showed that there is a direct and significant relationship between accident proneness and the personality traits of neuroticism and openness to experience. Furthermore, there is a meaningful but inverse correlation between accident proneness and the personality traits of extraversion and conscientiousness, while there was no relationship between accident proneness and the personality trait of agreeableness.

  20. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 2. Accident scenario based on thermodynamic model

    International Nuclear Information System (INIS)

    Maruyama, Shigenao

    2012-01-01

    An accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 2 is analyzed from the data open to the public. Phase equilibrium process model was introduced that the vapor and water are at saturation point in the vessels. Proposed accident scenario agrees very well with the data of the plant parameters obtained just after the accident. The estimation describes that the rupture time of the reactor pressure vessel (RPV) was at 22:50 14/3/2011. The estimation shows that the rupture time of the pressure containment vessel (RCP) was at 7:40 15/3/2011. These estimations are different from the ones by TEPCO, however; many measured evidences show good accordance with the present scenario. (author)

  1. Methodological development for selection of significant predictors explaining fatal road accidents.

    Science.gov (United States)

    Dadashova, Bahar; Arenas-Ramírez, Blanca; Mira-McWilliams, José; Aparicio-Izquierdo, Francisco

    2016-05-01

    Identification of the most relevant factors for explaining road accident occurrence is an important issue in road safety research, particularly for future decision-making processes in transport policy. However model selection for this particular purpose is still an ongoing research. In this paper we propose a methodological development for model selection which addresses both explanatory variable and adequate model selection issues. A variable selection procedure, TIM (two-input model) method is carried out by combining neural network design and statistical approaches. The error structure of the fitted model is assumed to follow an autoregressive process. All models are estimated using Markov Chain Monte Carlo method where the model parameters are assigned non-informative prior distributions. The final model is built using the results of the variable selection. For the application of the proposed methodology the number of fatal accidents in Spain during 2000-2011 was used. This indicator has experienced the maximum reduction internationally during the indicated years thus making it an interesting time series from a road safety policy perspective. Hence the identification of the variables that have affected this reduction is of particular interest for future decision making. The results of the variable selection process show that the selected variables are main subjects of road safety policy measures. Published by Elsevier Ltd.

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

    International Nuclear Information System (INIS)

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

    2011-08-01

    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

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

    International Nuclear Information System (INIS)

    Cheong, Jae Hak

    1993-02-01

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

  4. A Time Series Model for Assessing the Trend and Forecasting the Road Traffic Accident Mortality.

    Science.gov (United States)

    Yousefzadeh-Chabok, Shahrokh; Ranjbar-Taklimie, Fatemeh; Malekpouri, Reza; Razzaghi, Alireza

    2016-09-01

    Road traffic accident (RTA) is one of the main causes of trauma and known as a growing public health concern worldwide, especially in developing countries. Assessing the trend of fatalities in the past years and forecasting it enables us to make the appropriate planning for prevention and control. This study aimed to assess the trend of RTAs and forecast it in the next years by using time series modeling. In this historical analytical study, the RTA mortalities in Zanjan Province, Iran, were evaluated during 2007 - 2013. The time series analyses including Box-Jenkins models were used to assess the trend of accident fatalities in previous years and forecast it for the next 4 years. The mean age of the victims was 37.22 years (SD = 20.01). From a total of 2571 deaths, 77.5% (n = 1992) were males and 22.5% (n = 579) were females. The study models showed a descending trend of fatalities in the study years. The SARIMA (1, 1, 3) (0, 1, 0) 12 model was recognized as a best fit model in forecasting the trend of fatalities. Forecasting model also showed a descending trend of traffic accident mortalities in the next 4 years. There was a decreasing trend in the study and the future years. It seems that implementation of some interventions in the recent decade has had a positive effect on the decline of RTA fatalities. Nevertheless, there is still a need to pay more attention in order to prevent the occurrence and the mortalities related to traffic accidents.

  5. A Time Series Model for Assessing the Trend and Forecasting the Road Traffic Accident Mortality

    Science.gov (United States)

    Yousefzadeh-Chabok, Shahrokh; Ranjbar-Taklimie, Fatemeh; Malekpouri, Reza; Razzaghi, Alireza

    2016-01-01

    Background Road traffic accident (RTA) is one of the main causes of trauma and known as a growing public health concern worldwide, especially in developing countries. Assessing the trend of fatalities in the past years and forecasting it enables us to make the appropriate planning for prevention and control. Objectives This study aimed to assess the trend of RTAs and forecast it in the next years by using time series modeling. Materials and Methods In this historical analytical study, the RTA mortalities in Zanjan Province, Iran, were evaluated during 2007 - 2013. The time series analyses including Box-Jenkins models were used to assess the trend of accident fatalities in previous years and forecast it for the next 4 years. Results The mean age of the victims was 37.22 years (SD = 20.01). From a total of 2571 deaths, 77.5% (n = 1992) were males and 22.5% (n = 579) were females. The study models showed a descending trend of fatalities in the study years. The SARIMA (1, 1, 3) (0, 1, 0) 12 model was recognized as a best fit model in forecasting the trend of fatalities. Forecasting model also showed a descending trend of traffic accident mortalities in the next 4 years. Conclusions There was a decreasing trend in the study and the future years. It seems that implementation of some interventions in the recent decade has had a positive effect on the decline of RTA fatalities. Nevertheless, there is still a need to pay more attention in order to prevent the occurrence and the mortalities related to traffic accidents. PMID:27800467

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

    International Nuclear Information System (INIS)

    1992-01-01

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

  7. Investigating the Differences of Single-Vehicle and Multivehicle Accident Probability Using Mixed Logit Model

    Directory of Open Access Journals (Sweden)

    Bowen Dong

    2018-01-01

    Full Text Available Road traffic accidents are believed to be associated with not only road geometric feature and traffic characteristic, but also weather condition. To address these safety issues, it is of paramount importance to understand how these factors affect the occurrences of the crashes. Existing studies have suggested that the mechanisms of single-vehicle (SV accidents and multivehicle (MV accidents can be very different. Few studies were conducted to examine the difference of SV and MV accident probability by addressing unobserved heterogeneity at the same time. To investigate the different contributing factors on SV and MV, a mixed logit model is employed using disaggregated data with the response variable categorized as no accidents, SV accidents, and MV accidents. The results indicate that, in addition to speed gap, length of segment, and wet road surfaces which are significant for both SV and MV accidents, most of other variables are significant only for MV accidents. Traffic, road, and surface characteristics are main influence factors of SV and MV accident possibility. Hourly traffic volume, inside shoulder width, and wet road surface are found to produce statistically significant random parameters. Their effects on the possibility of SV and MV accident vary across different road segments.

  8. Development of MPS Method for Analyzing Melt Spreading Behavior and MCCI in Severe Accidents

    Science.gov (United States)

    Yamaji, Akifumi; Li, Xin

    2016-08-01

    Spreading of molten core (corium) on reactor containment vessel floor and molten corium-concrete interaction (MCCI) are important phenomena in the late phase of a severe accident for assessment of the containment integrity and managing the severe accident. The severe accident research at Waseda University has been advancing to show that simulations with moving particle semi-implicit (MPS) method (one of the particle methods) can greatly improve the analytical capability and mechanical understanding of the melt behavior in severe accidents. MPS models have been developed and verified regarding calculations of radiation and thermal field, solid-liquid phase transition, buoyancy, and temperature dependency of viscosity to simulate phenomena, such as spreading of corium, ablation of concrete by the corium, crust formation and cooling of the corium by top flooding. Validations have been conducted against experiments such as FARO L26S, ECOKATS-V1, Theofanous, and SPREAD for spreading, SURC-2, SURC-4, SWISS-1, and SWISS-2 for MCCI. These validations cover melt spreading behaviors and MCCI by mixture of molten oxides (including prototypic UO2-ZrO2), metals, and water. Generally, the analytical results show good agreement with the experiment with respect to the leading edge of spreading melt and ablation front history of concrete. The MPS results indicate that crust formation may play important roles in melt spreading and MCCI. There is a need to develop a code for two dimensional MCCI experiment simulation with MPS method as future study, which will be able to simulate anisotropic ablation of concrete.

  9. Health effects models for off-site radiological consequence analysis of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Togawa, Orihiko; Homma, Toshimitsu; Matsuzuru, Hideo; Kobayashi, Sadayoshi

    1991-02-01

    A first version of models has been developed for predicting the number of occurrences of health effects induced by radiation exposure in nuclear reactor accidents. The models are based on the health effects models developed originally by Harvard University (NUREG/CR-4214). These models are revised on the basis of the new information on risk estimates by the reassessment of the radiation dosimetry in Hiroshima and Nagasaki. The models deal with the following effects: (1) early effects models for bone marrow, lungs, gastrointestinal tract, central nervous system, thyroid, skin and reproductive organs, using the Weibull function, (2) late somatic effects models including leukemia and cancers of breast, lungs, thyroid, gastrointestinal tract and so forth, on the basis of the information derived from epidemiological studies on the atomic bomb survivors of Hiroshima and Nagasaki, (3) models for late and developmental effects due to exposure in utero. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1992-04-15

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

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

    International Nuclear Information System (INIS)

    1992-04-01

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

  12. Model of cyclist accident characteristics in the city of Malang and Blitar

    Science.gov (United States)

    Arifin, M. Z.; Agustin, I. W.

    2018-01-01

    Utilization of bicycles as an environmentally friendly mode of transportation is reconcerned as the development of sustainable transportation programs. The use of bicycles in some developed countries such as the Netherlands is 27 per cent of total travel, while for developing countries such as Indonesia, cyclists are less than 1% of total travel with low educated characteristics (65 per cent) and low income (48 per cent). Cyclist reduces dependencies on petroleum and environmental pollution as well as lessen the occurrence of congestion and traffic accidents involving motor vehicles. It was necessary to know the behavior and interaction of bicycle riders with other vehicle users in a heterogeneous traffic flow. The main purpose of the research is to create a model of bicycle accidents to increase the road traffic safety in Malang city and Blitar city. The research used analyses of frequency, the road’s level of service, and multiple linear regression. The results showed that there was a need for a development basis of a special lane for bicycle. It aims to reduce the level and number of cyclist accidents and to achieve transportation safety as well as to raise public awareness in traffic safety.

  13. Simulation Modeling Requirements for Loss-of-Control Accident Prevention of Turboprop Transport Aircraft

    Science.gov (United States)

    Crider, Dennis; Foster, John V.

    2012-01-01

    In-flight loss of control remains the leading contributor to aviation accident fatalities, with stall upsets being the leading causal factor. The February 12, 2009. Colgan Air, Inc., Continental Express flight 3407 accident outside Buffalo, New York, brought this issue to the forefront of public consciousness and resulted in recommendations from the National Transportation Safety Board to conduct training that incorporates stalls that are fully developed and develop simulator standards to support such training. In 2010, Congress responded to this accident with Public Law 11-216 (Section 208), which mandates full stall training for Part 121 flight operations. Efforts are currently in progress to develop recommendations on implementation of stall training for airline pilots. The International Committee on Aviation Training in Extended Envelopes (ICATEE) is currently defining simulator fidelity standards that will be necessary for effective stall training. These recommendations will apply to all civil transport aircraft including straight-wing turboprop aircraft. Government-funded research over the previous decade provides a strong foundation for stall/post-stall simulation for swept-wing, conventional tail jets to respond to this mandate, but turboprops present additional and unique modeling challenges. First among these challenges is the effect of power, which can provide enhanced flow attachment behind the propellers. Furthermore, turboprops tend to operate for longer periods in an environment more susceptible to ice. As a result, there have been a significant number of turboprop accidents as a result of the early (lower angle of attack) stalls in icing. The vulnerability of turboprop configurations to icing has led to studies on ice accumulation and the resulting effects on flight behavior. Piloted simulations of these effects have highlighted the important training needs for recognition and mitigation of icing effects, including the reduction of stall margins

  14. Simulation on Poisson and negative binomial models of count road accident modeling

    Science.gov (United States)

    Sapuan, M. S.; Razali, A. M.; Zamzuri, Z. H.; Ibrahim, K.

    2016-11-01

    Accident count data have often been shown to have overdispersion. On the other hand, the data might contain zero count (excess zeros). The simulation study was conducted to create a scenarios which an accident happen in T-junction with the assumption the dependent variables of generated data follows certain distribution namely Poisson and negative binomial distribution with different sample size of n=30 to n=500. The study objective was accomplished by fitting Poisson regression, negative binomial regression and Hurdle negative binomial model to the simulated data. The model validation was compared and the simulation result shows for each different sample size, not all model fit the data nicely even though the data generated from its own distribution especially when the sample size is larger. Furthermore, the larger sample size indicates that more zeros accident count in the dataset.

  15. Chernobyl and Fukushima nuclear accidents: what has changed in the use of atmospheric dispersion modeling?

    Science.gov (United States)

    Benamrane, Y; Wybo, J-L; Armand, P

    2013-12-01

    The threat of a major accidental or deliberate event that would lead to hazardous materials emission in the atmosphere is a great cause of concern to societies. This is due to the potential large scale of casualties and damages that could result from the release of explosive, flammable or toxic gases from industrial plants or transport accidents, radioactive material from nuclear power plants (NPPs), and chemical, biological, radiological or nuclear (CBRN) terrorist attacks. In order to respond efficiently to such events, emergency services and authorities resort to appropriate planning and organizational patterns. This paper focuses on the use of atmospheric dispersion modeling (ADM) as a support tool for emergency planning and response, to assess the propagation of the hazardous cloud and thereby, take adequate counter measures. This paper intends to illustrate the noticeable evolution in the operational use of ADM tools over 25 y and especially in emergency situations. This study is based on data available in scientific publications and exemplified using the two most severe nuclear accidents: Chernobyl (1986) and Fukushima (2011). It appears that during the Chernobyl accident, ADM were used few days after the beginning of the accident mainly in a diagnosis approach trying to reconstruct what happened, whereas 25 y later, ADM was also used during the first days and weeks of the Fukushima accident to anticipate the potentially threatened areas. We argue that the recent developments in ADM tools play an increasing role in emergencies and crises management, by supporting stakeholders in anticipating, monitoring and assessing post-event damages. However, despite technological evolutions, its prognostic and diagnostic use in emergency situations still arise many issues. Copyright © 2013 Elsevier Ltd. All rights reserved.

  16. Development of Northeast Asia Nuclear Power Plant Accident Simulator.

    Science.gov (United States)

    Kim, Juyub; Kim, Juyoul; Po, Li-Chi Cliff

    2017-06-15

    A conclusion from the lessons learned after the March 2011 Fukushima Daiichi accident was that Korea needs a tool to estimate consequences from a major accident that could occur at a nuclear power plant located in a neighboring country. This paper describes a suite of computer-based codes to be used by Korea's nuclear emergency response staff for training and potentially operational support in Korea's national emergency preparedness and response program. The systems of codes, Northeast Asia Nuclear Accident Simulator (NANAS), consist of three modules: source-term estimation, atmospheric dispersion prediction and dose assessment. To quickly assess potential doses to the public in Korea, NANAS includes specific reactor data from the nuclear power plants in China, Japan and Taiwan. The completed simulator is demonstrated using data for a hypothetical release. © The Author 2016. Published by Oxford University Press. All rights reserved. For Permissions, please email: journals.permissions@oup.com.

  17. Development of a Methodology for VHTR Accident Consequence Assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

    2016-05-15

    The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

  18. A drug cost model for injuries due to road traffic accidents.

    Directory of Open Access Journals (Sweden)

    Riewpaiboon A

    2008-03-01

    Full Text Available Objective: This study aimed to develop a drug cost model for injuries due to road traffic accidents for patients receiving treatment at a regional hospital in Thailand. Methods: The study was designed as a retrospective, descriptive analysis. The cases were all from road traffic accidents receiving treatment at a public regional hospital in the fiscal year 2004. Results: Three thousand seven hundred and twenty-three road accident patients were included in the study. The mean drug cost per case was USD18.20 (SD=73.49, median=2.36. The fitted drug cost model had an adjusted R2 of 0.449. The positive significant predictor variables of drug costs were prolonged length of stay, age over 30 years old, male, Universal Health Coverage Scheme, time of accident during 18:00-24:00 o’clock, and motorcycle comparing to bus. To forecast the drug budget for 2006, there were two approaches identified, the mean drug cost and the predicted average drug cost. The predicted average drug cost was calculated based on the forecasted values of statistically significant (p<0.05 predictor variables included in the fitted model; predicted total drug cost was USD44,334. Alternatively, based on the mean cost, predicted total drug cost in 2006 was USD63,408. This was 43% higher than the figure based on the predicted cost approach.Conclusions: The planned budget of drug cost based on the mean cost and predicted average cost were meaningfully different. The application of a predicted average cost model could result in a more accurate budget planning than that of a mean statistic approach.

  19. HYSPLIT's Capability for Radiological Aerial Monitoring in Nuclear Emergencies: Model Validation and Assessment on the Chernobyl Accident

    International Nuclear Information System (INIS)

    Jung, Gunhyo; Kim, Juyoul; Shin, Hyeongki

    2007-01-01

    The Chernobyl accident took place on 25 April 1986 in Ukraine. Consequently large amount of radionuclides were released into the atmosphere. The release was a widespread distribution of radioactivity throughout the northern hemisphere, mainly across Europe. A total of 31 persons died as a consequence of the accident, and about 140 persons suffered various degrees of radiation sickness and health impairment in the acute health impact. The possible increase of cancer incidence has been a real and significant increase of carcinomas of the thyroid among the children living in the contaminated regions as the late health effects. Recently, a variety of atmospheric dispersion models have been developed and used around the world. Among them, HYSPLIT (HYbrid Single-Particle Lagrangian Integrated Trajectory) model developed by NOAA (National Oceanic and Atmospheric Administration)/ARL (Air Resources Laboratory) is being widely used. To verify the HYSPLIT model for radiological aerial monitoring in nuclear emergencies, a case study on the Chernobyl accident is performed

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

    International Nuclear Information System (INIS)

    Son, Donggun; Kim, Dong-Ha; Park, Rae-Jun; Bae, Jun-Ho; Shim, Suk-Ku; Marigomen, Ralph

    2014-01-01

    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

  1. Development Status of Accident Tolerant Fuels for Light Water Reactors in Korea

    Energy Technology Data Exchange (ETDEWEB)

    Yang, Jae Ho; Kim, Hyun Gil; In, Wang Kee; Kim, Weon Ju; Koo, Yang Hyum [KAERI, Daejeon (Korea, Republic of); Lee, Seung Jae [KEPCONF, Daejeon (Korea, Republic of)

    2016-05-15

    Research on accident tolerant fuels (ATFs) is aimed at developing innovative fuels, which can mitigate or prevent the consequences of accidents. In Korea, innovative concepts are being developed to improve fuel safety and reliability of LWRs during accident events and normal operations. ATF technologies will be developed and commercialized through a sequence of long-lead and extensive activities. The interim milestone for new fuel program is that we would be ready for an irradiation test in commercial reactor by 2021. This presentation deals with the status of ATF development in KOREA and plan to implement new fuel technology successfully in commercial nuclear power plants.

  2. Development of analytical model for condensation of vapor mixture of nitric acid and water affected volatilized ruthenium behavior in accident of evaporation to dryness by boiling of reprocessed high level liquid waste at fuel reprocessing facilities

    International Nuclear Information System (INIS)

    Yoshida, Kazuo

    2016-08-01

    An accident of evaporation to dryness by boiling of high level liquid waste is postulated as one of the severe accidents caused by the loss of cooling function at a fuel reprocessing plant. In this case, continuous vaporing of nitric acid and water leads to increase Ru volatilization in liquid waste temperature over 120degC at later boiling and dry out phases. It has been observed at the experiments with actual and synthetic liquid waste that some amount of Ru volatilizes and transfers into condensed nitric acid solution at those phases. The nitric acid and water vapor flowing from waste tank are expected to condense at compartments of actual facilities building. The volatilized Ru could transfer into condensed liquid. It is key issues for quantifying the amount of transferred Ru through the facility building to simulate these thermodynamic and chemical behaviors. An analytical model has been proposed in this report based on the condensation mechanisms of nitric acid and water in vapor-liquid equilibria. It has been also carried out for the proposed model being feasible to formulate the activity coefficients and to review the thermodynamic properties of nitric acid solution. Practicability of the proposed analytical model has been shown successfully through the feasibility study with simulation of an experiment result. (author)

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

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

  4. Fuel models and results from the TRAC-PF1/MIMAS TMI-2 accident calculation

    International Nuclear Information System (INIS)

    Schwegler, E.C.; Maudlin, P.J.

    1983-01-01

    A brief description of several fuel models used in the TRAC-PF1/MIMAS analysis of the TMI-2 accident is presented, and some of the significant fuel-rod behavior results from this analysis are given. Peak fuel-rod temperatures, oxidation heat production, and embrittlement and failure behavior calculated for the TMI-2 accident are discussed. Other aspects of fuel behavior, such as cladding ballooning and fuel-cladding eutectic formation, were found not to significantly affect the accident progression

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

    International Nuclear Information System (INIS)

    Chang, Hyun Sop

    1996-02-01

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

  6. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    Energy Technology Data Exchange (ETDEWEB)

    Fritsching, G.; Wolf, G. [Internationale Atomreaktorbau G.m.b.H. (INTERATOM), Bergisch Gladbach (Germany, F.R.)

    1981-01-15

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen.

  7. Model experiments on depressurisation accidents in nuclear process heat plants (HTGR)

    International Nuclear Information System (INIS)

    Fritsching, G.; Wolf, G.

    1981-01-01

    The analysis of depressurisation accidents requires the use of digital computer programs to find out the dynamic loads acting on the plant structures. Because of the importance of such accidents in safety and licensing procedures of nuclear process heat plants, it is necessary to compare these computer results with suitable experiments to show the accuracy and the limits of the programs in question. For this purpose a series of depressurisation experiments has been started at INTERATOM on a small scale model of a primary loop of a nuclear process heat plant. Using the results of these experiments three different computer programs were tested with good success. The development of the experimental program and the estimation of the results was carried out in co-operation with KFA-Juelich and the Technische Hochschule Aachen

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

    International Nuclear Information System (INIS)

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

    2000-05-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-05-01

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

  10. The causing model of accidents and preventing system of small mines

    Energy Technology Data Exchange (ETDEWEB)

    Cao, S.; Zhang, L.; Liu, Y.; Li, Y. [Chongqing University, Chongqing (China)

    2008-06-15

    From an analysis of data on fatal accidents in small coal mines in a southern region of China over a period of three years, the time and type of accidents was discussed by applying statistical methods. It is shown that accidents frequently occur at the end of spring and all through summer. Roof accidents and gas disasters constitute severe accidents and traffic accidents are also important. It was found that most accidents are caused by dangerous behaviour of personnel and the unsafe state of equipment combined with economic interest. The three-factor causing model (TFC model) was proposed. Unsafe behaviour is a direct cause influenced by staff and workers while the unsafe nature of equipment is an indirect cause of accidents influence by natural conditions and the level of technical equipment in the mines. A system of accident prevention in small coal collieries was established with the TFC model. In this, scientific management is an important factor. 13 refs., 4 figs., 1 tab.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2007-07-01

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

  12. Developing an external domino accident prevention framework : Hazwim

    NARCIS (Netherlands)

    Reniers, Genserik L L; Dullaert, W.; Ale, B. J.M.; Soudan, K.

    Empirical research on major accident safety in the second largest chemical cluster worldwide, the Antwerp port area, supports the design of a meta-technical framework for optimizing external domino prevention. First, the majority of Seveso top tier companies have expressed a willingness to cooperate

  13. A Comparative analysis for control rod drop accident in RETRAN DNB and CETOP DNB Model

    International Nuclear Information System (INIS)

    Yang, Chang Keun; Kim, Yo Han; Ha, Sang Jun

    2009-01-01

    In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering, for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may occur difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor(PWR) has been raised. Due to the above reason, the Korea Electric Power Research Institute(KEPRI) had decided to develop the new safety analysis code system for Korea Standard Nuclear Power Plants in Korea. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the investigation for few candidates, RETRAN-3D has been chosen as a system analysis code. As a part of the feasibility estimation for the methodology and code system, CRD(Control Rod Drop) accident which an event of Non-LOCA accidents for Uljin units 3 and 4 and Yonggwang 1 and 2 was selected to verify the feasibility of the methodology using the RETRAN-3D. In this paper, RETRAN DNB Model and CETOP DNB Model were analyzed by using comparative method

  14. Modeling of in-vessel fission product release including fuel morphology effects for severe accident analyses

    International Nuclear Information System (INIS)

    Suh, K.Y.

    1989-10-01

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO 2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

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

    International Nuclear Information System (INIS)

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

    2015-01-01

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

  16. Analyses of systems availability and operator actions to support the development of severe accident procedures

    International Nuclear Information System (INIS)

    Lutz, R.J. Jr.; Scobel, J.H.

    1989-01-01

    This paper reports on traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, that have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severer accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses indicate that the traditional analyses result in overfly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. Additionally, since the traditional analyses do not model all of the operator actions which are prescribed, the impact of additional severe accident operator actions on the progression and consequences of the accident cannot be reliably identified. Further, the more detailed analysis can change the focus of the importance of various system to the prevention of core damage and the mitigation of severe accident consequences. Finally, the simplicity of the traditional analyses can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

  17. Strategies of modeling the cognitive tasks of human operators for accident scenarios in nuclear power plant control rooms

    International Nuclear Information System (INIS)

    Cheon, Se Woo; Sur, Sang Moon; Lee, Yong Hee; Lee, Jeong Wun

    1993-01-01

    This paper presents the development strategies of cognitive task network modeling for accident scenarios in nuclear power plant control rooms. Task network modeling is used to provide useful predictions of operator's performance times and error rates, based upon plant procedures and/or control room changes. Two accident scenarios, small-break loss of coolant accident (LOCA) and steam generator tube rupture (SGTR), are selected for task simulation. To obtain the input data for the model, task elements are extracted by the task analysis of emergency operating procedures. The input data include task performance time, communication ink, panel location, component operating mode, and data for performance shaping factors (PSFs). Operator's verbs are categorized according to the elements of cognitive behavior. The simulation of the task network for the small-break LOCA scenario is presented in this paper. (Author)

  18. Study on Developments in Accident Investigation Methods: A Survey of the 'State-of-the-Art'

    Energy Technology Data Exchange (ETDEWEB)

    Hollnagel, Erik; Speziali, Josephine (Ecole des Mines de Paris, F-06904 Sophia Antipolis (France))

    2008-01-15

    The objective of this project was to survey the main accident investigation methods that have been developed since the early or mid-1990s. The motivation was the increasing frequency of accidents that defy explanations in simple terms, for instance cause-effect chains or 'human error'. Whereas the complexity of socio-technical systems is steadily growing across all industrial domains, including nuclear power production, accident investigation methods are only updated when their inability to account for novel types of accidents and incidents becomes inescapable. Accident investigation methods therefore typically lag behind the socio-technological developments by 20 years or more. The project first compiled a set of methods from the recognised scientific literature and in major major research and development programs, excluding methods limited to risk assessment, technological malfunctions, human reliability, and safety management methods. An initial set of 21 methods was further reduced to seven by retaining only prima facie accident investigation methods and avoiding overlapping or highly similar methods. The second step was to develop a set of criteria used to characterise the methods. The starting point was Perrow's description of normal accidents in socio-technical systems, which used the dimensions of coupling, going from loose to tight, and interactions, going from linear to complex. For practical reasons, the second dimension was changed to that of tractability or how easy it is to describe the system, where the sub-criteria are the level of detail, the availability of an articulated model, and the system dynamics. On this basis the seven selected methods were characterised in terms of the systems - or conditions - they could account for, leading to the following four groups: methods suitable for systems that are loosely coupled and tractable, methods suitable for systems that are tightly coupled and tractable, methods suitable for systems that

  19. Radiological health effects models for nuclear power plant accident consequence analysis

    International Nuclear Information System (INIS)

    Evans, J.S.; Moeller, D.W.

    1989-01-01

    Improved health effects models have been developed for assessing the early effects, late somatic effects and genetic effects that might result from low-LET radiation exposures to populations following a major accident in a nuclear power plant. All the models have been developed in such a way that the dynamics of population risks can be analyzed. Estimates of life years lost and the duration of illnesses were generated and a framework recommended for summarizing health impacts. Uncertainty is addressed by providing models for upper, central and lower estimates of most effects. The models are believed to be a significant improvement over the models used in the U.S. Nuclear Regulatory Commission's Reactor Safety Study, and they can easily be modified to reflect advances in scientific understanding of the health effects of ionizing radiation

  20. A Model for Traffic Accidents Prediction Based on Driver Personality Traits Assessment

    Directory of Open Access Journals (Sweden)

    Marjana Čubranić-Dobrodolac

    2017-12-01

    Full Text Available The model proposed in this paper uses four psychological instruments for assessing driver behaviour and personality traits aiming to find a relationship between the considered constructs and the occurrence of traffic accidents. A Barratt Impulsiveness Scale (BIS-11 was used for the assessment of impulsivity, Aggressive Driving Behaviour Questionnaire (ADBQ for assessing the aggressiveness while driving, Manchester Driver Attitude Questionnaire (DAQ and the Questionnaire for self-assessment of driving ability. Besides these instruments, the participants filled out an extensive demographic survey. Within the statistical analysis, in addition to the descriptive indicators, correlation coefficients were calculated and four hierarchical regression analyses were performed to determine the predictive power of personality traits on the occurrence of traffic accidents. Further, to confirm the results and to obtain additional information about the relationship between the considered variables, the structural equation modelling and binary logistic regression have been implemented. A sample of this research covered 305 drivers, of which there were 100 bus drivers and 102 truck drivers, as well as 103 drivers of privately owned vehicles. The results indicate that BIS-11 and ADBQ questionnaires show the best predictive power which means that impulsivity and aggressiveness as personality traits have the greatest influence on the occurrence of traffic accidents. This research could be useful in many fields, such as the design of selection procedures for professional drivers, development of programs for the prevention of traffic accidents and violations of law, rehabilitation of drivers who have been deprived of the driving license, etc.

  1. Compartment model for long-term contamination prediction in deciduous fruit trees after a nuclear accident

    International Nuclear Information System (INIS)

    Antonopoulos-Domis, M.; Clouvas, A.; Gagianas, A.

    1990-01-01

    Radiocesium contamination from the Chernobyl accident of different parts (fruits, leaves, and shoots) of selected apricot trees in North Greece was systematically measured in 1987 and 1988. The results are presented and discussed in the framework of a simple compartment model describing the long-term contamination uptake mechanism of deciduous fruit trees after a nuclear accident

  2. Risk/benefits of fast breeder reactors societal risks and accident modelling

    International Nuclear Information System (INIS)

    Farmer, F.R.

    1979-01-01

    A special feature of the Fast Breeder Reactor is the possibility of fuel vaporisation, hence accidents may have more severe consequences than thermal reactors. This article discusses the process of accident modelling, the identification and assessment of risk, not yet incurred. 10 refs

  3. Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents

    International Nuclear Information System (INIS)

    Chang, Y.H.J.; Mosleh, A.

    2007-01-01

    This is the third in a series of five papers describing the IDAC (Information, Decision, and Action in Crew context) model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model is developed to probabilistically predict the responses of the nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper discusses the modeling components and their process rules. An operator's problem-solving process is divided into three types: information pre-processing (I), diagnosis and decision-making (D), and action execution (A). Explicit and context-dependent behavior rules for each type of operator are developed in the form of tables, and logical or mathematical relations. These regulate the process and activities of each of the three types of response. The behavior rules are developed for three generic types of operator: Decision Maker, Action Taker, and Consultant. This paper also provides a simple approach to calculating normalized probabilities of alternative behaviors given a context

  4. US nuclear industry approach to severe accident management guidance development and implementation

    International Nuclear Information System (INIS)

    Modeen, D.; Walsh, L.; Oehlberg, R.

    1992-01-01

    The purpose of this paper is to discuss the US nuclear industry activities, occurring under the auspices of Nuclear Management and Resources Council (NUMARC), to define, develop and implement enhancements to utility accident management capabilities. This effort consists of three major parts: (1) Development of a practical framework for evaluation of plant-specific accident management capabilities and the subsequent implementation of selected enhancements. (2) Development of specific technical guidance that address arresting core damage if it begins, either in-vessel or ex-vessel, and maintaining containment integrity. Preventing inadequate core cooling or minimizing the consequences of offsite releases, while considered to be candidate areas for accident management enhancements, have been the subject of intense previous study and development. (3) Plant-specific implementation of accident management enhancements in three areas: (a) personnel resources (organization, training, communications); (b) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (c) information resources (procedures and guidance, technical information, process information)

  5. U.S. nuclear industry approach to severe accident management guidance, development and implementation

    International Nuclear Information System (INIS)

    Modeen, D.; Walsh, L.; Oehlberg, R.

    1991-01-01

    The purpose of this paper is to discuss the US nuclear industry activities, occurring under the auspices of Nuclear Management and Resources Council (NUMARC), to define, develop and implement enhancements to utility accident management capabilities. This effort consists of three major parts: (1) Development of a practical framework for evaluation of plant-specific accident management capabilities and the subsequent implementation of selected enhancements. (2) Development of specific technical guidance that address arresting core damage if it begins, either in-vessel or ex-vessel, and maintaining containment integrity. Preventing inadequate core cooling or minimizing the consequences of offsite releases, while considered to be candidate areas for accident management enhancements, have been the subject of intense previous study and development. (3) Plant-specific implementation of accident management enhancements in three areas: (a) personnel resources (organization, training, communications); (b) systems and equipment (restoration and repair, instrumentation, use of alternatives); and (c) information resources (procedures and guidance, technical information, process information)

  6. Development of auditing technology for accident analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Kim, H. C.; Bae, K. H.; Lee, Y. J.; Chung, Y. J.; Jeong, J. J. [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    2003-06-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. The current year fall under the first step of the 3 year project, and the main researches were focused on identifying the candidate thermal hydraulic models for improvement. Well known PIRT methodology has been applied to identify model improvement items. As a part of PIRT process, the identification of SMART-P system and compenent has been performed. The scenario of each key accident and phenonema have been identified. To identify SMART-P thermal-hydraulic characteristics, preliminary calculation has been performed and identify the applicability and inprovement items of current auditing code, RELAP5.

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

    International Nuclear Information System (INIS)

    Lecomte, C.

    1990-06-01

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

  8. A Complex Network Model for Analyzing Railway Accidents Based on the Maximal Information Coefficient

    International Nuclear Information System (INIS)

    Shao Fu-Bo; Li Ke-Ping

    2016-01-01

    It is an important issue to identify important influencing factors in railway accident analysis. In this paper, employing the good measure of dependence for two-variable relationships, the maximal information coefficient (MIC), which can capture a wide range of associations, a complex network model for railway accident analysis is designed in which nodes denote factors of railway accidents and edges are generated between two factors of which MIC values are larger than or equal to the dependent criterion. The variety of network structure is studied. As the increasing of the dependent criterion, the network becomes to an approximate scale-free network. Moreover, employing the proposed network, important influencing factors are identified. And we find that the annual track density-gross tonnage factor is an important factor which is a cut vertex when the dependent criterion is equal to 0.3. From the network, it is found that the railway development is unbalanced for different states which is consistent with the fact. (paper)

  9. Modelling of radioactive fallout in the vicinity of Chernobyl nuclear power plant accident

    International Nuclear Information System (INIS)

    Israel, Y.A.; Petrov, V.N.; Severov, D.A.

    1988-03-01

    Deposition of radioactive products escaping into the atmosphere for a long time from the Chernobylsk-4 reactor resulting in residual radioactive contamination of the region at a distance of up to 100 km from the nuclear power plant is considered. The suggested model may be used for estimation of the possible scope of nuclear danger in the regions of nuclear power plants and creation of conditions ensuring safety of the population at possible accidents. The following topics are developed: height of elevation and conditions of radionuclide transfer in the atmosphere; dynamics of release and dispersive composition of radioactive products; calculations of radiation levels at a close trace [fr

  10. A simplified model for calculating early offsite consequences from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1988-07-01

    A personal computer-based model, SMART, has been developed that uses an integral approach for calculating early offsite consequences from nuclear reactor accidents. The solution procedure uses simplified meteorology and involves direct analytic integration of air concentration equations over time and position. This is different from the discretization approach currently used in the CRAC2 and MACCS codes. The SMART code is fast-running, thereby providing a valuable tool for sensitivity and uncertainty studies. The code was benchmarked against both MACCS version 1.4 and CRAC2. Results of benchmarking and detailed sensitivity/uncertainty analyses using SMART are presented. 34 refs., 21 figs., 24 tabs

  11. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    International Nuclear Information System (INIS)

    Yoshida, Yoshitaka; Ohtani, Masanori; Fujita, Yushi

    2002-01-01

    In the nuclear power plant, much knowledge is acquired through probabilistic safety assessment (PSA) of a severe accident, and accident management (AM) is prepared. It is necessary to evaluate the effectiveness of AM using the decision-making failure probability of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipments in PSA. However, there has been no suitable qualification method for PSA so far to obtain the decision-making failure probability, because the decision-making failure of an emergency organization treats the knowledge based error. In this work, we developed a new method for quantification of the decision-making failure probability of an emergency organization using cognitive analysis model, which decided an AM strategy, in a nuclear power plant at the severe accident, and tried to apply it to a typical pressurized water reactor (PWR) plant. As a result: (1) It could quantify the decision-making failure probability adjusted to PSA for general analysts, who do not necessarily possess professional human factors knowledge, by choosing the suitable value of a basic failure probability and an error-factor. (2) The decision-making failure probabilities of six AMs were in the range of 0.23 to 0.41 using the screening evaluation method and in the range of 0.10 to 0.19 using the detailed evaluation method as the result of trial evaluation based on severe accident analysis of a typical PWR plant, and a result of sensitivity analysis of the conservative assumption, failure probability decreased about 50%. (3) The failure probability using the screening evaluation method exceeded that using detailed evaluation method by 99% of probability theoretically, and the failure probability of AM in this study exceeded 100%. From this result, it was shown that the decision-making failure probability was more conservative than the detailed evaluation method, and the screening evaluation method satisfied

  12. Quantification of a decision-making failure probability of the accident management using cognitive analysis model

    Energy Technology Data Exchange (ETDEWEB)

    Yoshida, Yoshitaka; Ohtani, Masanori [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan); Fujita, Yushi [TECNOVA Corp., Tokyo (Japan)

    2002-09-01

    In the nuclear power plant, much knowledge is acquired through probabilistic safety assessment (PSA) of a severe accident, and accident management (AM) is prepared. It is necessary to evaluate the effectiveness of AM using the decision-making failure probability of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipments in PSA. However, there has been no suitable qualification method for PSA so far to obtain the decision-making failure probability, because the decision-making failure of an emergency organization treats the knowledge based error. In this work, we developed a new method for quantification of the decision-making failure probability of an emergency organization using cognitive analysis model, which decided an AM strategy, in a nuclear power plant at the severe accident, and tried to apply it to a typical pressurized water reactor (PWR) plant. As a result: (1) It could quantify the decision-making failure probability adjusted to PSA for general analysts, who do not necessarily possess professional human factors knowledge, by choosing the suitable value of a basic failure probability and an error-factor. (2) The decision-making failure probabilities of six AMs were in the range of 0.23 to 0.41 using the screening evaluation method and in the range of 0.10 to 0.19 using the detailed evaluation method as the result of trial evaluation based on severe accident analysis of a typical PWR plant, and a result of sensitivity analysis of the conservative assumption, failure probability decreased about 50%. (3) The failure probability using the screening evaluation method exceeded that using detailed evaluation method by 99% of probability theoretically, and the failure probability of AM in this study exceeded 100%. From this result, it was shown that the decision-making failure probability was more conservative than the detailed evaluation method, and the screening evaluation method satisfied

  13. Development of Safety Significance Evaluation Program for Accidents and Events in NPPs

    International Nuclear Information System (INIS)

    Yang, Hui Chang; Hong, Seok Jin; Cho, Nam Chul; Chung, Dae Wook; Lee, Chang Joo

    2010-01-01

    To evaluate the significance in terms of safety for the accidents and events occurred in nuclear power plants using probabilistic safety assessment techniques can provide useful insights to the regulator. Based on the quantified risk information of accident or event occurred, regulators can decide which regulatory areas should be focused than the others. To support these regulatory analysis activities, KINS-ASP program was developed. KINS-ASP program can supports the risk increase due to the occurred accidents or events by providing the graphic interfaces and linked quantification engines for the PSA experts and non- PSA acquainted regulators both

  14. Development of Auditing Technology for Accident Analysis of SMART-P

    Energy Technology Data Exchange (ETDEWEB)

    Chung, B. D.; Lee, Y. J.; Jeong, J. J.; Kim, H. C.; Chung, Y. J.; Bae, K. H

    2006-02-15

    The objective of this project is to develop thermal hydraulic models of the regulatory auditing codes for the application of SMART-P integrated reactor. At initial period, PIRT has been performed to identify the model deficiencies and determine the priority of model improvements. The identified thermal hydraulic models has been implemented to RELAP5/MOD3.3 auditing code according to the PIRT ranking. The input model for SMART-P has been developed with consistent to the current design status documents and checked by independent reviewer as Q/A procedure.The evaluation of experimental availabilities and code collapsible has been done by expert group and summarized as validation matrix forms. The experimental data of VISTA, which is the only integral effect test facility, were used to validate the improved model. The safety analysis has been demonstrated for the essential accident scenario. The validation and demonstration show that the developed models are applicable to utilize in reliable and independent auditing for SMART design certification.

  15. Installation places of criticality accident detectors in the plutonium conversion development facility

    International Nuclear Information System (INIS)

    Sanada, Yukihisa; Tsujimura, Norio; Shimizu, Yoshio; Izaki, Kenji; Furuta, Sadaaki

    2008-01-01

    At the Plutonium Conversion Development Facility (PCDF) in the Nuclear Fuel Cycle Engineering Laboratories, the co-conversion technologies to purify the mixed plutonium and uranium nitrate solution discharged from a reprocessing plant have been developed. The probability of a criticality accident in PCDF is extremely low. However, the criticality accident alarm system (CAAS) has been in place since 1982 to reduce the radiation dose to workers in case of such a rare criticality accident. The CAAS contains criticality accident detector units (CADs), one unit consisting of three plastic scintillation detectors, and using the 2 out of 3 voting system for the purpose of high reliability. Currently, eight CADs are installed in PCDF evaluating the dose using a simple equation allowing for a safety margin. The purpose of this study is to show the determination procedures for the adequate relocation of the CADs which adequately ensures safety in PCDF. (author)

  16. Estimation Of 137Cs Using Atmospheric Dispersion Models After A Nuclear Reactor Accident

    Science.gov (United States)

    Simsek, V.; Kindap, T.; Unal, A.; Pozzoli, L.; Karaca, M.

    2012-04-01

    Nuclear energy will continue to have an important role in the production of electricity in the world as the need of energy grows up. But the safety of power plants will always be a question mark for people because of the accidents happened in the past. Chernobyl nuclear reactor accident which happened in 26 April 1986 was the biggest nuclear accident ever. Because of explosion and fire large quantities of radioactive material was released to the atmosphere. The release of the radioactive particles because of accident affected not only its region but the entire Northern hemisphere. But much of the radioactive material was spread over west USSR and Europe. There are many studies about distribution of radioactive particles and the deposition of radionuclides all over Europe. But this was not true for Turkey especially for the deposition of radionuclides released after Chernobyl nuclear reactor accident and the radiation doses received by people. The aim of this study is to determine the radiation doses received by people living in Turkish territory after Chernobyl nuclear reactor accident and use this method in case of an emergency. For this purpose The Weather Research and Forecasting (WRF) Model was used to simulate meteorological conditions after the accident. The results of WRF which were for the 12 days after accident were used as input data for the HYSPLIT model. NOAA-ARL's (National Oceanic and Atmospheric Administration Air Resources Laboratory) dispersion model HYSPLIT was used to simulate the 137Cs distrubition. The deposition values of 137Cs in our domain after Chernobyl Nuclear Reactor Accident were between 1.2E-37 Bq/m2 and 3.5E+08 Bq/m2. The results showed that Turkey was affected because of the accident especially the Black Sea Region. And the doses were calculated by using GENII-LIN which is multipurpose health physics code.

  17. Multidisciplinary response for a radiological accident developed in Argentina

    International Nuclear Information System (INIS)

    Di Giorgio, Marina; Tadic, Matilde; Vazquez, Marina; Cateriano, Miguel; Jordan, Osvaldo; Rojas, Carlos A.; Portas, M.; Coppola, A.; De Lellis, M. del C.; Dovasio, F.

    2010-01-01

    In January 2009 during an oil well logging task in Rincon de los Sauces, Province of Neuquen, Argentina, and, as a result of the temporary loss of control of a 137 Cs sealed source of 92,5 GBq (2,5 Ci), two workers were allegedly exposed to ionizing radiation. The Health Emergency Coordination System in Neuquen activated the Radiological Emergency Intervention System of the Nuclear Regulatory Authority (NRA). An initial assessment of the accident scenario and the involved persons was performed. Two workers were were transfer to Buenos Aires for a biodosimetry evaluation. In Buenos Aires, the Burn Hospital performed the first diagnostic and therapeutic approach of the radioinduced skin injuries. The NRA estimated the biological absorbed doses and the studies of telethermography. The Italian Hospital was in charge of the studies using ultrasonography and Eco- Doppler. As a result of the multidisciplinary evaluation performed, one of the workers was diagnosed with superficial radioinduced dermatitis. The mentioned intervention highlighted the importance of multidisciplinary and multiparametric work for the evaluation of a radiological accident, as well as for the diagnosis and early treatment of the exposed persons. (authors) [es

  18. The impact of safety design consideration on future LMFBR developments. (R and D needs related to accident accommodation)

    International Nuclear Information System (INIS)

    Justin, F.

    1985-04-01

    Accident accommodation for design accidents or even beyond design basis accidents is based on components and systems for which important research and development work is needed. Main issues are treated: fuel failure faults, sodium fires, decay heat removal, accommodation of energetics and debris

  19. Integral thermal model of severe accident dynamics of NPP with containment

    International Nuclear Information System (INIS)

    Arutyunyan, R.V.; Bol'shov, L.A.; Vasil'ev, A.D.; Kamennov, G.P.

    1991-01-01

    An analytical model of the interaction of reactor core remains with concrete during severe accidents at nuclear power plants is considered. Time dependences of side and radial concrete melting are plotted. Time dependences of containment atmosphere temperature and pressure during a severe accident at nuclear power plants are investigated analytically and numerically. The sensitivity of the results to the coefficient values in the problem is studied within the range of their concertainty. The Kaverna-1 is described. The results of modelling a severe NPP accident which have been obtained using the Kaverna-1 package are presented

  20. Interactive Rapid Dose Assessment Model (IRDAM): reactor-accident assessment methods. Vol.2

    International Nuclear Information System (INIS)

    Poeton, R.W.; Moeller, M.P.; Laughlin, G.J.; Desrosiers, A.E.

    1983-05-01

    As part of the continuing emphasis on emergency preparedness, the US Nuclear Regulatory Commission (NRC) sponsored the development of a rapid dose assessment system by Pacific Northwest Laboratory (PNL). This system, the Interactive Rapid Dose Assessment Model (IRDAM) is a micro-computer based program for rapidly assessing the radiological impact of accidents at nuclear power plants. This document describes the technical bases for IRDAM including methods, models and assumptions used in calculations. IRDAM calculates whole body (5-cm depth) and infant thyroid doses at six fixed downwind distances between 500 and 20,000 meters. Radionuclides considered primarily consist of noble gases and radioiodines. In order to provide a rapid assessment capability consistent with the capacity of the Osborne-1 computer, certain simplifying approximations and assumptions are made. These are described, along with default values (assumptions used in the absence of specific input) in the text of this document. Two companion volumes to this one provide additional information on IRDAM. The user's Guide (NUREG/CR-3012, Volume 1) describes the setup and operation of equipment necessary to run IRDAM. Scenarios for Comparing Dose Assessment Models (NUREG/CR-3012, Volume 3) provides the results of calculations made by IRDAM and other models for specific accident scenarios

  1. A simplified model for calculating atmospheric radionuclide transport and early health effects from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Madni, I.K.; Cazzoli, E.G.; Khatib-Rahbar, M.

    1995-01-01

    During certain hypothetical severe accidents in a nuclear power plant, radionuclides could be released to the environment as a plume. Prediction of the atmospheric dispersion and transport of these radionuclides is important for assessment of the risk to the public from such accidents. A simplified PC-based model was developed that predicts time-integrated air concentration of each radionuclide at any location from release as a function of time integrated source strength using the Gaussian plume model. The solution procedure involves direct analytic integration of air concentration equations over time and position, using simplified meteorology. The formulation allows for dry and wet deposition, radioactive decay and daughter buildup, reactor building wake effects, the inversion lid effect, plume rise due to buoyancy or momentum, release duration, and grass height. Based on air and ground concentrations of the radionuclides, the early dose to an individual is calculated via cloudshine, groundshine, and inhalation. The model also calculates early health effects based on the doses. This paper presents aspects of the model that would be of interest to the prediction of environmental flows and their public consequences

  2. A model for the calculation of the off-site economic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Gallego, E.; Alonso, A.

    1988-01-01

    The off-site economic cost of nuclear reactor accidents will depend on the countermeasures adopted to reduce its radiological impact. The assessment of the direct costs of emergency countermeasures (evacuation, early relocation and food disposal) as well as those of long-term protective actions (food disposal, decontamination or interdiction) is the objective of a model under development, with the sponsorship of the CEC Radiation Protection Programme, called MECA (Model for assessing the Economic Consequences of Accidents). The meteorological and socio-economical peculiarities of each site studied will be taken into account, by means of a flexible meteorological sampling scheme, which considers the geographical distribution of population and economic centers, and a data-base, compatible with the existing European grid, that contains the population distribution and the economic characteristics of the environs of the site to be studied with more detail near the reactor. The paper summarizes the particular models which will be included in MECA and shows the importance of site-specific adaptable modelling for economic risk evaluation

  3. A model national emergency plan for radiological accidents; Plan modelo nacional de respuesta de emergencia para accidentes radiologicos

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2000-07-01

    The IAEA has supported several projects for the development of a national response plan for radiological emergencies. As a result, the IAEA has developed a model National Emergency Response Plan for Radiological Accidents (RAD PLAN), particularly for countries that have no nuclear power plants. This plan can be adapted for use by countries interested in developing their own national radiological emergency response plan, and the IAEA will supply the latest version of the RAD PLAN on computer diskette upon request.

  4. Modeling requirements for full-scope reactor simulators of fission-product transport during severe accidents

    International Nuclear Information System (INIS)

    Ellison, P.G.; Monson, P.R.; Mitchell, H.A.

    1990-01-01

    This paper describes in the needs and requirements to properly and efficiently model fission product transport on full scope reactor simulators. Current LWR simulators can be easily adapted to model severe accident phenomena and the transport of radionuclides. Once adapted these simulators can be used as a training tool during operator training exercises for training on severe accident guidelines, for training on containment venting procedures, or as training tool during site wide emergency training exercises

  5. Advanced surrogate model and sensitivity analysis methods for sodium fast reactor accident assessment

    International Nuclear Information System (INIS)

    Marrel, A.; Marie, N.; De Lozzo, M.

    2015-01-01

    Within the framework of the generation IV Sodium Fast Reactors, the safety in case of severe accidents is assessed. From this statement, CEA has developed a new physical tool to model the accident initiated by the Total Instantaneous Blockage (TIB) of a sub-assembly. This TIB simulator depends on many uncertain input parameters. This paper aims at proposing a global methodology combining several advanced statistical techniques in order to perform a global sensitivity analysis of this TIB simulator. The objective is to identify the most influential uncertain inputs for the various TIB outputs involved in the safety analysis. The proposed statistical methodology combining several advanced statistical techniques enables to take into account the constraints on the TIB simulator outputs (positivity constraints) and to deal simultaneously with various outputs. To do this, a space-filling design is used and the corresponding TIB model simulations are performed. Based on this learning sample, an efficient constrained Gaussian process metamodel is fitted on each TIB model outputs. Then, using the metamodels, classical sensitivity analyses are made for each TIB output. Multivariate global sensitivity analyses based on aggregated indices are also performed, providing additional valuable information. Main conclusions on the influence of each uncertain input are derived. - Highlights: • Physical-statistical tool for Sodium Fast Reactors TIB accident. • 27 uncertain parameters (core state, lack of physical knowledge) are highlighted. • Constrained Gaussian process efficiently predicts TIB outputs (safety criteria). • Multivariate sensitivity analyses reveal that three inputs are mainly influential. • The type of corium propagation (thermal or hydrodynamic) is the most influential

  6. Development on quantitative safety analysis method of accident scenario. The automatic scenario generator development for event sequence construction of accident

    International Nuclear Information System (INIS)

    Kojima, Shigeo; Onoue, Akira; Kawai, Katsunori

    1998-01-01

    This study intends to develop a more sophisticated tool that will advance the current event tree method used in all PSA, and to focus on non-catastrophic events, specifically a non-core melt sequence scenario not included in an ordinary PSA. In the non-catastrophic event PSA, it is necessary to consider various end states and failure combinations for the purpose of multiple scenario construction. Therefore it is anticipated that an analysis work should be reduced and automated method and tool is required. A scenario generator that can automatically handle scenario construction logic and generate the enormous size of sequences logically identified by state-of-the-art methodology was developed. To fulfill the scenario generation as a technical tool, a simulation model associated with AI technique and graphical interface, was introduced. The AI simulation model in this study was verified for the feasibility of its capability to evaluate actual systems. In this feasibility study, a spurious SI signal was selected to test the model's applicability. As a result, the basic capability of the scenario generator could be demonstrated and important scenarios were generated. The human interface with a system and its operation, as well as time dependent factors and their quantification in scenario modeling, was added utilizing human scenario generator concept. Then the feasibility of an improved scenario generator was tested for actual use. Automatic scenario generation with a certain level of credibility, was achieved by this study. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1997-07-01

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

  8. Modelling of plate-out under gas-cooled reactor (GCR) accident conditions

    International Nuclear Information System (INIS)

    Taig, A.R.

    1981-01-01

    The importance of plate-out in mitigating consequences of gas-cooled reactor accidents, and its place in assessing these consequences, are discussed. The data requirements of a plate-out modelling program are discussed, and a brief description is given of parallel work programs on thermal/hydraulic reactor behaviour and fuel modelling, both of which will provide inputs to the plate-out program under development. The representation of a GCR system used in SRD studies is presented, and the equations governing iodine adsorption, desorption and transport round the circuit are derived. The status of SRD's plate-out program is described, and the type of sensitivity studies to be undertaken with the partially-developed computer program in order to identify the most useful lines for future research is discussed. (author)

  9. A source term estimation method for a nuclear accident using atmospheric dispersion models

    DEFF Research Database (Denmark)

    Kim, Minsik; Ohba, Ryohji; Oura, Masamichi

    2015-01-01

    The objective of this study is to develop an operational source term estimation (STE) method applicable for a nuclear accident like the incident that occurred at the Fukushima Dai-ichi nuclear power station in 2011. The new STE method presented here is based on data from atmospheric dispersion...... models and short-range observational data around the nuclear power plants.The accuracy of this method is validated with data from a wind tunnel study that involved a tracer gas release from a scaled model experiment at Tokai Daini nuclear power station in Japan. We then use the methodology developed...... and validated through the effort described in this manuscript to estimate the release rate of radioactive material from the Fukushima Dai-ichi nuclear power station....

  10. Modelling of RPV lower head under core melt severe accident condition using OpenFOAM

    International Nuclear Information System (INIS)

    Madokoro, Hiroshi; Kretzschmar, Frank; Miassoedov, Alexei

    2017-01-01

    Although six years have been passed since the tragic severe accident at Fukushima Daiichi, still large uncertainties exist in modeling of core degradation and reactor pressure vessel (RPV) failure. It is extremely important to obtain a better understanding of complex phenomena in the lower head in order to improve accident management measures. The possible failure mode of reactor pressure vessel and its failure time are especially a matter of importance. Thermal behavior of the molten pool can be simulated by the Phase-change Effective Convectivity Model (PECM), which is a distributed-parameter model developed in the Royal Institute of Technology (KTH), Sweden. The model calculates convective currents not using a pure CFD approach but based on so called “characteristic velocities” that are determined by empirical correlations depending on the geometry and physical properties of the molten pool. At the Karlsruhe Institute of Technology (KIT), the PECM has been implemented in the open-source CFD software OpenFOAM in order to receive detailed predictions of a core melt behavior in the RPV lower head under severe accident conditions. An advantage of using OpenFOAM is that it is very flexible to add and modify models and physical properties. In the current work, the solver is extended to couple PECM with a structure analysis model of the vessel wall. The model considers thermal expansion, plasticity, creep and damage. The model and physical properties are based on those implemented in ANSYS. Although the previous implementation had restriction that the amount of and geometry of the melt cannot be changed, our coupled model allows flexibility of the melt amount and geometry. The extended solver was used to simulate the LIVE-L1 and -L7V experiments and has demonstrated good prediction of the temperature distribution in the molten pool and heat flux distribution through the vessel wall. Regarding the vessel failure the model was applied to one of the FOREVER tests

  11. Research and development with regard to severe accidents in pressurised water reactors: Summary and outlook

    International Nuclear Information System (INIS)

    2011-01-01

    transport in the primary and the secondary circuit (7.2), aerosol behaviour in the containment (7.4) and FP chemistry (7.5). Finally, Chapter 8 presents a review of development and validation efforts for the main severe accident codes: ASTEC, MAAP and MELCOR. In Chapters 3-7, for each of the theme areas, the phenomena involved are reviewed. The major relevant experiments are then briefly described, including recent, ongoing and future projects. The key models and specific codes (except for integral codes) used to simulate the phenomena in question are also discussed. Finally, the state of current knowledge is reviewed and an outlook for the future is presented, especially regarding experimental programmes and the development of modelling tools. (authors)

  12. Experimental and modelling studies of iodine oxide formation and aerosol behaviour relevant to nuclear reactor accidents

    International Nuclear Information System (INIS)

    Dickinson, S.; Auvinen, A.; Ammar, Y.; Bosland, L.; Clément, B.; Funke, F.; Glowa, G.; Kärkelä, T.; Powers, D.A.; Tietze, S.; Weber, G.; Zhang, S.

    2014-01-01

    Highlights: • Radiolytic reactions can influence iodine volatility following a nuclear accident. • Kinetic models have been developed based on atmospheric chemistry studies. • Properties of iodine oxide aerosols produced by radiation have been measured. • Decomposition of iodine oxides by the action of heat or radiation has been observed. - Abstract: Plant assessments have shown that iodine contributes significantly to the source term for a range of accident scenarios. Iodine has a complex chemistry that determines its chemical form and, consequently, its volatility in the containment. If volatile iodine species are formed by reactions in the containment, they will be subject to radiolytic reactions in the atmosphere, resulting in the conversion of the gaseous species into involatile iodine oxides, which may deposit on surfaces or re-dissolve in water pools. The concentration of airborne iodine in the containment will, therefore, be determined by the balance between the reactions contributing to the formation and destruction of volatile species, as well as by the physico-chemical properties of the iodine oxide aerosols which will influence their longevity in the atmosphere. This paper summarises the work that has been done in the framework of the EC SARNET (Severe Accident Research Network) to develop a greater understanding of the reactions of gaseous iodine species in irradiated air/steam atmospheres, and the nature and behaviour of the reaction products. This work has mainly been focussed on investigating the nature and behaviour of iodine oxide aerosols, but earlier work by members of the SARNET group on gaseous reaction rates is also discussed to place the more recent work into context

  13. Modeling and measuring the effects of imprecision in accident management

    International Nuclear Information System (INIS)

    Yu, Donghan

    2002-01-01

    This paper presents two approaches for evaluating the uncertainties inherent in accident management strategies. Current PRA methodology uses expert opinion in the assessment of rare event probabilities. The problem is that these probabilities may be difficult to estimate even though reasonable engineering judgement is applied. This occurs because expert opinion under incomplete knowledge and limited data is inherently imprecise. In this case, the concept of uncertainty about a probability value is both intuitively appealing and potentially useful. This analysis considers accident management as a decision problem (i.e. 'applying a strategy' vs. 'do nothing') and uses an influence diagram. Then, the analysis applies two approaches to evaluating imprecise node probabilities in the influence diagram: 'a fuzzy probability' and 'an interval-valued subjective probability'. For the propagation of subjective probabilities, the analysis uses a Monte-Carlo simulation approach. In case of fuzzy probabilities, fuzzy logic is applied to propagate them. We believe that these approaches can allow us to better understand uncertainties associated with severe accident management strategies, because they provide additional information regarding the implications of using imprecise input data

  14. Joint research project WASA-BOSS: Further development and application of severe accident codes. Assessment and optimization of accident management measures. Project B: Accident analyses for pressurized water reactors with the application of the ATHLET-CD code

    International Nuclear Information System (INIS)

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

    2017-02-01

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

  15. Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

    International Nuclear Information System (INIS)

    Oh, Jae Yong; Yun, Jong Il; Kim, Do Sam; Han Chul

    2011-01-01

    Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

  16. Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

    Energy Technology Data Exchange (ETDEWEB)

    Oh, Jae Yong; Yun, Jong Il [KAIST, Daejeon (Korea, Republic of); Kim, Do Sam; Han Chul [Korea Institue of Nuclear Safety, Daejeon (Korea, Republic of)

    2011-05-15

    Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

  17. A dynamic food-chain model and program for predicting the radiological consequences of nuclear accident

    International Nuclear Information System (INIS)

    Hu Erbang; Gao Zhanrong; Zhang Heyuan; Wei Weiqiang

    1996-12-01

    A dynamic food-chain model and program, DYFOM-95, for predicting the radiological consequences of nuclear accident has been developed, which is not only suitable to the West food-chain but also to Chinese food chain. The following processes, caused by accident release which will make an impact on radionuclide concentration in the edible parts of vegetable are considered: dry and wet deposition interception and initial retention, translocation, percolation, root uptake and tillage. Activity intake rate of animals, effects of processing and activity intake of human through ingestion pathway are also considered in calculations. The effects of leaf area index LAI of vegetable are considered in dry deposition model. A method for calculating the contribution of rain with different period and different intensity to total wet deposition is established. The program contains 1 main code and 5 sub-codes to calculate dry and wet deposition on surface of vegetable and soil, translocation of nuclides in vegetable, nuclide concentration in the edible parts of vegetable and in animal products and activity intake of human and so on. (24 refs., 9 figs., 11 tabs.)

  18. Validation of CFD modeling for VGM loss-of-forced-cooling accidents

    International Nuclear Information System (INIS)

    Wysocki, Aaron; Ahmed, Bobby; Charmeau, Anne; Anghaie, Samim

    2009-01-01

    Heat transfer and fluid flow in the VGM reactor cavity cooling system (RCCS) was modeled using Computational Fluid Dynamics (CFD). The VGM is a Russian modular-type high temperature helium-cooled reactor. In the reactor cavity, heat is removed from the pressure vessel wall through natural convection and radiative heat transfer to water-cooled vertical pipes lining the outer cavity concrete. The RCCS heat removal capability under normal operation and accident scenarios needs to be assessed. The purpose of the present study is to validate the use of CFD to model heat transfer in the VGM RCCS. Calculations were based on a benchmark problem which defines a two-dimensional temperature distribution on the pressure vessel outer wall for both Depressurized and Pressurized Loss-of-Forced Cooling events. A two-dimensional axisymmetric model was developed to determine the best numerical modeling approach. A grid sensitivity study for the air region showed that a 20 mm mesh size with a boundary layer giving a maximum y+ of 2.0 was optimal. Sensitivity analyses determined that the discrete ordinates radiative model, the k-omega turbulence model, and the ideal gas law gave the best combination for capturing radiation and natural circulation in the air cavity. A maximum RCCS pipe wall temperature of 62degC located 6 m from the top of the cavity was predicted. The model showed good agreement with previous results for both Pressurized and Depressurized Loss-of-Forced-Cooling accidents based on RCCS coolant outlet temperature, relative contributions of radiative and convective heat transfer, and RCCS heat load profiles. (author)

  19. ASTEC V2 severe accident integral code: Fission product modelling and validation

    International Nuclear Information System (INIS)

    Cantrel, L.; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-01-01

    One main goal of the severe accident integral code ASTEC V2, jointly developed since almost more than 15 years by IRSN and GRS, is to simulate the overall behaviour of fission products (FP) in a damaged nuclear facility. ASTEC applications are source term determinations, level 2 Probabilistic Safety Assessment (PSA2) studies including the determination of uncertainties, accident management studies and physical analyses of FP experiments to improve the understanding of the phenomenology. ASTEC is a modular code and models of a part of the phenomenology are implemented in each module: the release of FPs and structural materials from degraded fuel in the ELSA module; the transport through the reactor coolant system approximated as a sequence of control volumes in the SOPHAEROS module; and the radiochemistry inside the containment nuclear building in the IODE module. Three other modules, CPA, ISODOP and DOSE, allow respectively computing the deposition rate of aerosols inside the containment, the activities of the isotopes as a function of time, and the gaseous dose rate which is needed to model radiochemistry in the gaseous phase. In ELSA, release models are semi-mechanistic and have been validated for a wide range of experimental data, and noticeably for VERCORS experiments. For SOPHAEROS, the models can be divided into two parts: vapour phase phenomena and aerosol phase phenomena. For IODE, iodine and ruthenium chemistry are modelled based on a semi-mechanistic approach, these FPs can form some volatile species and are particularly important in terms of potential radiological consequences. The models in these 3 modules are based on a wide experimental database, resulting for a large part from international programmes, and they are considered at the state of the art of the R and D knowledge. This paper illustrates some FPs modelling capabilities of ASTEC and computed values are compared to some experimental results, which are parts of the validation matrix

  20. Modeling of spreading of the melted corium jet inside the pool of emergency heat removal during severe accidents at NPP

    Directory of Open Access Journals (Sweden)

    I. V. Kazachkov

    2012-03-01

    Full Text Available Important nuclear power safety problem in touch with modeling of melted corium jet spreading inside the coolant pool is considered in the paper. It appears by development of the passive protection systems against se-vere accidents. The non-linear mathematical developed model is presented for the jet under reactor vessel pool for one of the perspective passive protection systems and the results of its analysis and studies are given. The performed analysis and the results of the numerical simulation done on the base of the model have allowed estab-lishing the interesting behaviors of the system, which may be useful for the scientists, as well as the engineers-constructors of the passive protection systems against severe accidents.

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

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  2. VICTORIA: A mechanistic model of radionuclide behavior in the reactor coolant system under severe accident conditions. Revision 1

    Energy Technology Data Exchange (ETDEWEB)

    Heams, T J [Science Applications International Corp., Albuquerque, NM (United States); Williams, D A; Johns, N A; Mason, A [UKAEA, Winfrith, (England); Bixler, N E; Grimley, A J [Sandia National Labs., Albuquerque, NM (United States); Wheatley, C J [UKAEA, Culcheth (England); Dickson, L W [Atomic Energy of Canada Ltd., Chalk River, ON (Canada); Osborn-Lee, I [Oak Ridge National Lab., TN (United States); Domagala, P; Zawadzki, S; Rest, J [Argonne National Lab., IL (United States); Alexander, C A [Battelle, Columbus, OH (United States); Lee, R Y [Nuclear Regulatory Commission, Washington, DC (United States)

    1992-12-01

    The VICTORIA model of radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident is described. It has been developed by the USNRC to define the radionuclide phenomena and processes that must be considered in systems-level models used for integrated analyses of severe accident source terms. The VICTORIA code, based upon this model, predicts fission product release from the fuel, chemical reactions involving fission products, vapor and aerosol behavior, and fission product decay heating. Also included is a detailed description of how the model is implemented in VICTORIA, the numerical algorithms used, and the correlations and thermochemical data necessary for determining a solution. A description of the code structure, input and output, and a sample problem are provided.

  3. Development of a multi-criteria decision tool for remediation after a nuclear or radiological accident

    Energy Technology Data Exchange (ETDEWEB)

    Luca, Christiano de; Rochedo, Elaine R.R.; Ferreira, Nadya M.P.D., E-mail: christiano_luca@hotmail.com, E-mail: elainerochedo@gmail.com, E-mail: nadya@ime.eb.br [Instituto Militar de Engenharia (IME), Rio de Janeiro, RJ (Brazil)

    2013-07-01

    The review of accidents involving Nuclear Power Plants or facilities that use or process radioactive sources have raised issues related to the decision-making processes and to the procedures used to reestablish the normal living conditions in the affected areas. Due to the large complexity of the decision processes after accidents, a multi-criteria approach has been recommended to support the choice among the several procedures that may improve the environmental conditions. As part of the process of developing a multi-criteria decision support tool, a questionnaire was created to be fulfilled by experts to derive the relevance of the technical criteria to be considered in the model. At this stage, only the technical criteria related to radiation protection of the public will be focused; legal aspects, costs and public opinion, although relevant in the decision-making process, are beyond the scope of this work. The questionnaire contains 12 questions, each containing 5 degrees of importance. The answers are statically analyzed to generate a multiplicative factor to be included in the multicriteria model. To facilitate the process of distributing the questionnaire to the selected experts and then for a better processing and ordering of the information gathered, a program based on the Hypertext Preprocessor language (PHP) was created; this methodology has been chosen because of its compatibility and security in existing operating systems. The relevance rank showed the long-term dose reduction and the generation of wastes as the most relevant aspects to be considered in selecting remediation strategies for a contaminated area. (author)

  4. Development of a multi-criteria decision tool for remediation after a nuclear or radiological accident

    International Nuclear Information System (INIS)

    Luca, Christiano de; Rochedo, Elaine R.R.; Ferreira, Nadya M.P.D.

    2013-01-01

    The review of accidents involving Nuclear Power Plants or facilities that use or process radioactive sources have raised issues related to the decision-making processes and to the procedures used to reestablish the normal living conditions in the affected areas. Due to the large complexity of the decision processes after accidents, a multi-criteria approach has been recommended to support the choice among the several procedures that may improve the environmental conditions. As part of the process of developing a multi-criteria decision support tool, a questionnaire was created to be fulfilled by experts to derive the relevance of the technical criteria to be considered in the model. At this stage, only the technical criteria related to radiation protection of the public will be focused; legal aspects, costs and public opinion, although relevant in the decision-making process, are beyond the scope of this work. The questionnaire contains 12 questions, each containing 5 degrees of importance. The answers are statically analyzed to generate a multiplicative factor to be included in the multicriteria model. To facilitate the process of distributing the questionnaire to the selected experts and then for a better processing and ordering of the information gathered, a program based on the Hypertext Preprocessor language (PHP) was created; this methodology has been chosen because of its compatibility and security in existing operating systems. The relevance rank showed the long-term dose reduction and the generation of wastes as the most relevant aspects to be considered in selecting remediation strategies for a contaminated area. (author)

  5. Environmental accident and its treatment in a developing country: a case study on China.

    Science.gov (United States)

    Hou, Yu

    2012-08-01

    Along with their rapid progress, developing countries have had to deal with more environmental problems, which have been a cause for concern among policy makers and the public in general. This study cites two accidents that happened in China in 2006 that caused serious environmental problems in nearby communities and discusses the problems these accidents created and the resulting disputes among the concerned people. Pollution-causing accidents not only pose threats to the health of the victims but also give rise to environmental disputes that jeopardise national security and social stability. Conflicts normally ensue following a pollution-causing accident, which are more likely to happen within a development zone or industrial park. Few environmental conflicts in the past decades were resolved through litigation. Nevertheless, there are lapses in the regulatory system, which have to be addressed to ensure that the public's rights and interests are protected. Currently, reports on pollution-causing accidents are difficult to obtain and are often released very late. A majority of industrial firms operate without environmental clearance, thus highlighting the government's inefficiency in environmental management. It is about time that the Chinese government takes seriously the use of the Environmental Impact Assessment.

  6. Development of a DNBR evaluation method for the CEA ejection accident in SMART core

    Energy Technology Data Exchange (ETDEWEB)

    Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

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

    International Nuclear Information System (INIS)

    Lee, Youngseung; Kim, Hyeongtaek; Shin, Jungmin

    2013-01-01

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

  8. Radiation protection survey of research and development activities initiated after the Chernobyl accident. Review report

    International Nuclear Information System (INIS)

    Burkart, W.

    1989-01-01

    The compilation of research and development activities in the various fields of radiation protection in OECD Member countries which have been undertaken or planned specifically to address open questions arising from the Chernobyl reactor accident experience shows a potential for international cooperative arrangements and/or coordination between national programmes. Both the preliminary review of the answers, which only cover a part of the relevant activities in OECD Member countries, and a computerized literature search indicate that the multidisciplinarity of the research area under consideration will call for special efforts to efficiently implement new models and new quantitative findings from the different fields of activity to provide an improved basis for emergency management and risk assessment. Further improvements could also be achieved by efforts to initiate new activities to close gaps in the programmes under way, to enhance international cooperation, and to coordinate the evaluation of the results. This preliminary review of the answers of 17 Member countries to the questionnaire on research and development activities initiated after the Chernobyl accident is not sufficient as a basis for a balanced decision on those research areas most in need for international cooperation and coordination. It may however serve as a guide for the exploration of the potential for international cooperative arrangements and/or coordination between national programmes by the CRPPH. Even at this preliminary stage, several specific activities are proposed to the NEA/OECD by Member countries. Whole body counting and the intercomparison of national data bases on the behaviour of radionuclides in the environment did attract most calls for international cooperation sponsored by the NEA

  9. Generation of risk importance information from severe accident PSA model

    International Nuclear Information System (INIS)

    Seo, Mi Ro; Kim, Hyeong Taek; Moon, Chan Kook

    2012-01-01

    One of the important objects conducting Probabilistic Safety Assessment (PSA) is the relative evaluation of importance of the component or function that is greatly affected to the plant safety. This evaluation is performed by the importance assessment methods such as Risk Reduction Worth, Risk Achievement Worth, and Fuss el Vessley method from the aspect of core damage frequency (CDF). In the Level 1 PSA model, the importance of each component can be evaluated since the CDF is calculated by the combination of the branch probability of event tree and the component failure probability in the fault tree. But, the Level 2 PSA model in order to assess the containment integrity cannot evaluate the risk importance by the above methods because the model is consisted of 3 parts, plant damage status, containment event tree, and source term category. So, in the field that the Level 2 PSA risk importance information should be reflected, such as maintenance rule program, risk importance has been determined by the subjective judgment of the model developer. This study was performed in order to generate the risk importance information more objectively and systematically in the Level 2 PSA model, focused on the containment event tree in the domain PHWR Level 2 PSA model

  10. Developing GIS based decision-support tools for agricultural counter-measurements after radiation accident

    International Nuclear Information System (INIS)

    Kepka, Pavel; Prochazka, Jan; Brom, Jakub; Pecharova, Emilie

    2009-01-01

    There is a whole variety of possibilities proposed by EURANOS data sheets for agriculture, for mid-term and long-term counter-measures after contamination of crops by radiation. We have developed a set of supportive tools for decision-makers within the project 'Methods of evaluation of contaminated territory after radiation accident - the importance of structure and functioning of a land cover'. Our TM tools are based on ArcGIS platform and Python programming language. We have developed a simple model for estimating the current biomass of the polluted crops. Inputs for this model are: a shape file of land cover data, database table with customisable plant growth characteristics and shape file of polluted areas. The model provides a shape file data set of estimated amounts of biomass of selected crops per hectare for a given day. The results are helpful for better performing of the countermeasure 'Early removal of crops'. The total amount of polluted waste, logistic costs (transport of people and material; required time; other costs) could be estimated only with basic GIS tools. The number of days expected for the harvest can be also calculated and compared with the dose and half-lives of the contaminating radionuclides. This analysis could also lead to a 'Do nothing' decision, especially in case of radionuclides with short times of half-life. (author)

  11. System Response Analysis of Rod Ejection Accident for APR1400 Using KNAP Hot Spot Model

    International Nuclear Information System (INIS)

    Kim, Yo-Han; Ha, Sang-Jun; Jun, Hwang-Yong

    2006-01-01

    Korea Electric Power Research Institute (KEPRI) has been developed the non-loss-of-coolant accident (non- LOCA) analysis methodology, called as the Korea Non- LOCA Analysis Package (KNAP), for the typical Optimized Power Reactor 1000 (OPR1000) plants. Considering current licensing methodology conducted by ABB-CE, however, the KNAP could be applied to Advanced Power Reactor 1400 (APR1400) also. In spite of some difference in design concepts of two plant types, there is a close resemblance between their nuclear steam supply systems (NSSS). So, in this study, the rod ejection accident (REA) event was analyzed using KNAP hot spot model (HSM) for APR1400 to estimate the feasibility of the application and the results were compared with those given in APR1400 Standard Safety Analysis Report (SSAR), which were calculated using the CESEC-III and STRIKIN-II code of ABB-CE. Through the study, it was concluded that the KNAP could be applicable to APR1400 on the view point of REA

  12. Possibility of the development of a Serbian protection system against chemical accidents

    Directory of Open Access Journals (Sweden)

    Dejan R. Inđić

    2012-10-01

    Full Text Available The paper presents a draft of a system model for responding in case of chemical accidents in accordance with the current legislation regarding the environment protection, the structure and elements of the existing response system in case of chemical accidents, other works dealing with the issue as well as the prospects planned by those responsible for the environmental protection. The paper discuss the possibilities of different institutions and agencies of the Republic of Serbia to engage in specialized methods of cooperation and protection against chemical hazards in accordance with Article X of the Convention on the Prohibition of Chemical Weapons.

  13. Development of a site-wide accident management center for the Savannah River Site

    International Nuclear Information System (INIS)

    Heal, D.W.; Britt, T.E.

    1992-01-01

    In 1990, the Safety Analysis Group at the Savannah River Site (SRS) began development of an Accident Management program. The program was designed to provide a total system which would meet the Department of Energy (DOE) Safety Performance Criteria, in regard to severe accident management, in the most effective manner. This paper will present two significant changes in the current SRS Accident Management program which will be used to meet these expanded needs. The first and most significant change will be to expand the diversity of the groups involved in the Accident Management process. In the future, organizations such as Environmental Safety, Health ampersand Quality Assurance, Emergency Planning, Site Management, Human Factors, Risk Assessment, and many others will work as an integrated team to solve facility problems. Organizations such as Materials Technology, Equipment Engineering and many of the laboratories on site will be utilized as support groups to increase the technical capability for specific accident analyses. This phase of the program is currently being structured, and should be operational by January of 1993

  14. Fast dose assessment models, parameters and code under accident conditions for Qinshan Nuclear Power Plant

    International Nuclear Information System (INIS)

    Zhang, Z.Y.; Hu, E.B.; Meng, X.C.; Zhang, Y.; Yao, R.T.

    1993-01-01

    According to requirement of accident emergency plan for Qinshan Nuclear Power Plant, a Gaussian straight-line model was adopted for estimating radionuclide concentration in surface air. In addition, the effects of mountain body on atmospheric dispersion was considered. By combination of field atmospheric dispersion experiment and wind tunnel modeling test, necessary modifications have been done for some models and parameters. A computer code for assessment was written in Quick BASIC (V4.5) language. The radius of assessment region is 10 km and the code is applicable to early accident assessment. (1 tab.)

  15. Development of an Accident Reproduction Simulator System Using a Hemodialysis Extracorporeal Circulation System

    Science.gov (United States)

    Nishite, Yoshiaki; Takesawa, Shingo

    2016-01-01

    Background: Accidents that occur during dialysis treatment are notified to the medical staff via alarms raised by the dialysis apparatus. Similar to such real accidents, apparatus activation or accidents can be reproduced by simulating a treatment situation. An alarm that corresponds to such accidents can be utilized in the simulation model. Objectives: The aim of this study was to create an extracorporeal circulation system (hereinafter, the circulation system) for dialysis machines so that it sets off five types of alarms for: 1) decreased arterial pressure, 2) increased arterial pressure, 3) decreased venous pressure, 4) increased venous pressure, and 5) blood leakage, according to the five types of accidents chosen based on their frequency of occurrence and the degree of severity. Materials and Methods: In order to verify the alarm from the dialysis apparatus connected to the circulation system and the accident corresponding to it, an evaluation of the alarm for its reproducibility of an accident was performed under normal treatment circumstances. The method involved testing whether the dialysis apparatus raised the desired alarm from the moment of control of the circulation system, and measuring the time it took until the desired alarm was activated. This was tested on five main models from four dialyzer manufacturers that are currently used in Japan. Results: The results of the tests demonstrated successful activation of the alarms by the dialysis apparatus, which were appropriate for each of the five types of accidents. The time between the control of the circulatory system to the alarm signal was as follows, 1) venous pressure lower limit alarm: 7 seconds; 2) venous pressure lower limit: 8 seconds; 3) venous pressure upper limit: 7 seconds; 4) venous pressure lower limit alarm: 2 seconds; and 5) blood leakage alarm: 19 seconds. All alarms were set off in under 20 seconds. Conclusions: Thus, we can conclude that a simulator system using an extracorporeal

  16. Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool.

    Science.gov (United States)

    Courageot, Estelle; Sayah, Rima; Huet, Christelle

    2010-05-07

    Estimating the dose distribution in a victim's body is a relevant indicator in assessing biological damage from exposure in the event of a radiological accident caused by an external source. When the dose distribution is evaluated with a numerical anthropomorphic model, the posture and morphology of the victim have to be reproduced as realistically as possible. Several years ago, IRSN developed a specific software application, called the simulation of external source accident with medical images (SESAME), for the dosimetric reconstruction of radiological accidents by numerical simulation. This tool combines voxel geometry and the MCNP(X) Monte Carlo computer code for radiation-material interaction. This note presents a new functionality in this software that enables the modelling of a victim's posture and morphology based on non-uniform rational B-spline (NURBS) surfaces. The procedure for constructing the modified voxel phantoms is described, along with a numerical validation of this new functionality using a voxel phantom of the RANDO tissue-equivalent physical model.

  17. Using Immersive Virtual Reality to Reduce Work Accidents in Developing Countries.

    Science.gov (United States)

    Nedel, Luciana; de Souza, Vinicius Costa; Menin, Aline; Sebben, Lucia; Oliveira, Jackson; Faria, Frederico; Maciel, Anderson

    2016-01-01

    Thousands of people die or are injured in work accidents every year. Although the lack of safety equipment is one of the causes, especially in developing countries, behavioral issues caused by psychosocial factors are also to blame. This article introduces the use of immersive VR simulators to preventively reduce accidents in the workplace by detecting behavioral patterns that may lead to an increased predisposition to risk exposure. The system simulates day-to-day situations, analyzes user reactions, and classifies the behaviors according to four psychosocial groups. The results of a user study support the effectiveness of this approach.

  18. Development of on-site accident criteria for waste transfer casks

    International Nuclear Information System (INIS)

    Uldrich, E.D.

    1989-01-01

    Removal of radioactive waste must withstand the scrutiny of the public and various regulatory offices. Currently, there is no standard accident criteria or methodology for intra-site shipments at the Idaho National Engineering Laboratory (INEL). Since the radioactive waste transfer casks only carry material within the INEL site boundaries and are not used for normal over-the-road transport, the requirements of 10 CFR 71 Packaging and Transportation of Radioactive Material, do not provide suitable requirements for cask design or safety analyses. The objective is to develop realistically conservative accident scenarios consistent with the limited uses at the INEL for which the cask is approved

  19. Development of reactor accident diagnostic system DISKET using knowledge engineering technique

    International Nuclear Information System (INIS)

    Yokobayashi, Masao; Yoshida, Kazuo; Kohsaka, Atsuo; Yamamoto, Minoru.

    1986-01-01

    An accident diagnostic system DISKET has been developed to identify the cause and the type of an abnormal transient of a nuclear power plant. The system is based on the knowledge engineering (KE) and consists of an inference engine IERIAS and a knowledge base. The main features of DISKET are the following : (1) Time-varying characteristics of transients can be treated. (2) Knowledge base can be divided into several knowledge units to handle a lot of rules effectively. (3) Programming language UTILISP, which is a dialect of LISP, is used to manipulate symbolic data effectively. For the verification of DISKET, performance tests have been conducted for several types of accidents. The knowledge base used in the tests was generated from the data of various types of transients produced by a PWR plant simulator. The results of verification studies showed a good applicability of DISKET to reactor accident diagnosis. (author)

  20. Trends in state-level freight accident rates: An enhancement of risk factor development for RADTRAN

    International Nuclear Information System (INIS)

    Saricks, C.; Kvitek, T.

    1991-01-01

    Under the Nuclear Waste Policy Act, the Department of Energy's Office of Civilian Radioactive Waste Management (OCRWM) is concerned with understanding and managing risk as it applies to the shipment of spent commercial nuclear reactor fuel. Understanding risk in relation to mode and geography may provide opportunities to minimize radiological and non-radiological risks of transportation. To enhance such an understanding, a set of state-or waterway-specific accident, fatality, and injury rates (expressed as rates per shipment kilometer) by transportation mode and highway administrative class was developed, using publicly-available data bases. Adjustments made to accommodate miscoded or incomplete information in accident data are described, as well as the procedures for estimating state-level flow data. Results indicate that the shipping conditions under which spent fuel is likely to be transported should be less subject to accidents than the ''average'' shipment within mode. 10 refs., 3 tabs

  1. Combined Prediction Model of Death Toll for Road Traffic Accidents Based on Independent and Dependent Variables

    Directory of Open Access Journals (Sweden)

    Feng Zhong-xiang

    2014-01-01

    Full Text Available In order to build a combined model which can meet the variation rule of death toll data for road traffic accidents and can reflect the influence of multiple factors on traffic accidents and improve prediction accuracy for accidents, the Verhulst model was built based on the number of death tolls for road traffic accidents in China from 2002 to 2011; and car ownership, population, GDP, highway freight volume, highway passenger transportation volume, and highway mileage were chosen as the factors to build the death toll multivariate linear regression model. Then the two models were combined to be a combined prediction model which has weight coefficient. Shapley value method was applied to calculate the weight coefficient by assessing contributions. Finally, the combined model was used to recalculate the number of death tolls from 2002 to 2011, and the combined model was compared with the Verhulst and multivariate linear regression models. The results showed that the new model could not only characterize the death toll data characteristics but also quantify the degree of influence to the death toll by each influencing factor and had high accuracy as well as strong practicability.

  2. Combined prediction model of death toll for road traffic accidents based on independent and dependent variables.

    Science.gov (United States)

    Feng, Zhong-xiang; Lu, Shi-sheng; Zhang, Wei-hua; Zhang, Nan-nan

    2014-01-01

    In order to build a combined model which can meet the variation rule of death toll data for road traffic accidents and can reflect the influence of multiple factors on traffic accidents and improve prediction accuracy for accidents, the Verhulst model was built based on the number of death tolls for road traffic accidents in China from 2002 to 2011; and car ownership, population, GDP, highway freight volume, highway passenger transportation volume, and highway mileage were chosen as the factors to build the death toll multivariate linear regression model. Then the two models were combined to be a combined prediction model which has weight coefficient. Shapley value method was applied to calculate the weight coefficient by assessing contributions. Finally, the combined model was used to recalculate the number of death tolls from 2002 to 2011, and the combined model was compared with the Verhulst and multivariate linear regression models. The results showed that the new model could not only characterize the death toll data characteristics but also quantify the degree of influence to the death toll by each influencing factor and had high accuracy as well as strong practicability.

  3. Development of passive condensers for accident localization systems at nuclear power plants in the former USSR

    International Nuclear Information System (INIS)

    Kuznecov, M.V.

    1992-01-01

    The development is summarized of passive condensers for accident localization systems at nuclear power plants (with RBMK and WWER reactors) in the former USSR. Basic properties and criteria defining their availability are described, as are experimental tests and technical solution optimization results. (author) 2 fig

  4. Initial VHTR accident scenario classification: models and data.

    Energy Technology Data Exchange (ETDEWEB)

    Vilim, R. B.; Feldman, E. E.; Pointer, W. D.; Wei, T. Y. C.; Nuclear Engineering Division

    2005-09-30

    Nuclear systems codes are being prepared for use as computational tools for conducting performance/safety analyses of the Very High Temperature Reactor. The thermal-hydraulic codes are RELAP5/ATHENA for one-dimensional systems modeling and FLUENT and/or Star-CD for three-dimensional modeling. We describe a formal qualification framework, the development of Phenomena Identification and Ranking Tables (PIRTs), the initial filtering of the experiment databases, and a preliminary screening of these codes for use in the performance/safety analyses. In the second year of this project we focused on development of PIRTS. Two events that result in maximum fuel and vessel temperatures, the Pressurized Conduction Cooldown (PCC) event and the Depressurized Conduction Cooldown (DCC) event, were selected for PIRT generation. A third event that may result in significant thermal stresses, the Load Change event, is also selected for PIRT generation. Gas reactor design experience and engineering judgment were used to identify the important phenomena in the primary system for these events. Sensitivity calculations performed with the RELAP5 code were used as an aid to rank the phenomena in order of importance with respect to the approach of plant response to safety limits. The overall code qualification methodology was illustrated by focusing on the Reactor Cavity Cooling System (RCCS). The mixed convection mode of heat transfer and pressure drop is identified as an important phenomenon for Reactor Cavity Cooling System (RCCS) operation. Scaling studies showed that the mixed convection mode is likely to occur in the RCCS air duct during normal operation and during conduction cooldown events. The RELAP5/ATHENA code was found to not adequately treat the mixed convection regime. Readying the code will require adding models for the turbulent mixed convection regime while possibly performing new experiments for the laminar mixed convection regime. Candidate correlations for the turbulent

  5. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.

    1985-01-01

    SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

  6. Dynamic modeling of physical phenomena for probabilistic assessment of spent fuel accidents

    International Nuclear Information System (INIS)

    Benjamin, A.S.

    1997-01-01

    If there should be an accident involving drainage of all the water from a spent fuel pool, the fuel elements will heat up until the heat produced by radioactive decay is balanced by that removed by natural convection to air, thermal radiation, and other means. If the temperatures become high enough for the cladding or other materials to ignite due to rapid oxidation, then some of the fuel might melt, leading to an undesirable release of radioactive materials. The amount of melting is dependent upon the fuel loading configuration and its age, the oxidation and melting characteristics of the materials, and the potential effectiveness of recovery actions. The authors have developed methods for modeling the pertinent physical phenomena and integrating the results with a probabilistic treatment of the uncertainty distributions. The net result is a set of complementary cumulative distribution functions for the amount of fuel melted

  7. Control rod ejection analysis during a depressurization accident and the development of a rod-ejection-preventing device

    International Nuclear Information System (INIS)

    Mitake, S.; Itoh, K.; Fukushima, H.; Inoue, T.

    1982-01-01

    The control rods used for the experimental VHTR are suspended in the core by means of flexible steel cables and it is conceivable that an accidental rod ejection could occur due to a depressurization accident. The computer code AFLADE was developed in order to analyze the possibility of accidental rod ejection, and several studies were performed. The parametric study results showed that the adopted design condition for the VHTR core will not cause a rod ejection accident. In parallel with these accident analyses, a rod-ejection-preventing device was developed in preparation for a hypothetical accident, and its function was verified by the component tests

  8. Advanced evacuation model managed through fuzzy logic during an accident in LNG terminal

    Energy Technology Data Exchange (ETDEWEB)

    Stankovicj, Goran; Petelin, Stojan [Faculty for Maritime Studies and Transport, University of Ljubljana, Portorozh (Sierra Leone); others, and

    2014-07-01

    Evacuation of people located inside the enclosed area of an LNG terminal is a complex problem, especially considering that accidents involving LNG are potentially very hazardous. In order to create an evacuation model managed through fuzzy logic, extensive influence must be generated from safety analyses. A very important moment in the optimal functioning of an evacuation model is the creation of a database which incorporates all input indicators. The output result is the creation of a safety evacuation route which is active at the moment of the accident. (Author)

  9. Risk assessment model for nuclear accident emergency protection countermeasure based on fuzzy matter-element analysis

    International Nuclear Information System (INIS)

    Xin Jing; Tang Huaqing; Zhang Yinghua; Zhang Limin

    2009-01-01

    A risk assessment model of nuclear accident emergency protection countermeasure based on fuzzy matter-element analysis and Euclid approach degree is proposed in the paper. The weight of assessed index is determined by information entropy and the scoring by experts, which could not only make full use of the inherent information of the indexes adequately, but reduce subjective assumption in the course of assessment effectively. The applied result shows that it is reasonable that the model is adopted to make risk assessment for nuclear accident emergency protective countermeasure,and it could be a kind of effective analytical method and decision making basis to choose the optimum protection countermeasure. (authors)

  10. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

    International Nuclear Information System (INIS)

    Evans, J.S.

    1990-01-01

    This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, models are included for assessing the risks of several nonlethal early and continuing effects -- including prodromal vomiting and diarrhea, hypothyroidism and radiation thyroiditis, skin burns, reproductive effects, and pregnancy losses. Linear and linear-quadratic models are recommended for estimating cancer risks. Parameters are given for analyzing the risks of seven types of cancer in adults -- leukemia, bone, lung, breast, gastrointestinal, thyroid, and ''other.'' The category, ''other'' cancers, is intended to reflect the combined risks of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus and cervix. Models of childhood cancers due to in utero exposure are also developed. For most cancers, both incidence and mortality are addressed. The models of cancer risk are derived largely from information summarized in BEIR III -- with some adjustment to reflect more recent studies. 64 refs., 18 figs., 46 tabs

  11. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S. (Harvard Univ., Boston, MA (USA). School of Public Health)

    1990-01-01

    This report describes dose-response models intended to be used in estimating the radiological health effects of nuclear power plant accidents. Models of early and continuing effects, cancers and thyroid nodules, and genetic effects are provided. Weibull dose-response functions are recommended for evaluating the risks of early and continuing health effects. Three potentially lethal early effects -- the hematopoietic, pulmonary, and gastrointestinal syndromes -- are considered. In addition, models are included for assessing the risks of several nonlethal early and continuing effects -- including prodromal vomiting and diarrhea, hypothyroidism and radiation thyroiditis, skin burns, reproductive effects, and pregnancy losses. Linear and linear-quadratic models are recommended for estimating cancer risks. Parameters are given for analyzing the risks of seven types of cancer in adults -- leukemia, bone, lung, breast, gastrointestinal, thyroid, and other.'' The category, other'' cancers, is intended to reflect the combined risks of multiple myeloma, lymphoma, and cancers of the bladder, kidney, brain, ovary, uterus and cervix. Models of childhood cancers due to in utero exposure are also developed. For most cancers, both incidence and mortality are addressed. The models of cancer risk are derived largely from information summarized in BEIR III -- with some adjustment to reflect more recent studies. 64 refs., 18 figs., 46 tabs.

  12. Usefulness of high resolution coastal models for operational oil spill forecast: the "Full City" accident

    Directory of Open Access Journals (Sweden)

    G. Broström

    2011-11-01

    Full Text Available Oil spill modeling is considered to be an important part of a decision support system (DeSS for oil spill combatment and is useful for remedial action in case of accidents, as well as for designing the environmental monitoring system that is frequently set up after major accidents. Many accidents take place in coastal areas, implying that low resolution basin scale ocean models are of limited use for predicting the trajectories of an oil spill. In this study, we target the oil spill in connection with the "Full City" accident on the Norwegian south coast and compare operational simulations from three different oil spill models for the area. The result of the analysis is that all models do a satisfactory job. The "standard" operational model for the area is shown to have severe flaws, but by applying ocean forcing data of higher resolution (1.5 km resolution, the model system shows results that compare well with observations. The study also shows that an ensemble of results from the three different models is useful when predicting/analyzing oil spill in coastal areas.

  13. Developing a Minimum Data Set for an Information Management System to Study Traffic Accidents in Iran.

    Science.gov (United States)

    Mohammadi, Ali; Ahmadi, Maryam; Gharagozlu, Alireza

    2016-03-01

    Each year, around 1.2 million people die in the road traffic incidents. Reducing traffic accidents requires an exact understanding of the risk factors associated with traffic patterns and behaviors. Properly analyzing these factors calls for a comprehensive system for collecting and processing accident data. The aim of this study was to develop a minimum data set (MDS) for an information management system to study traffic accidents in Iran. This descriptive, cross-sectional study was performed in 2014. Data were collected from the traffic police, trauma centers, medical emergency centers, and via the internet. The investigated resources for this study were forms, databases, and documents retrieved from the internet. Forms and databases were identical, and one sample of each was evaluated. The related internet-sourced data were evaluated in their entirety. Data were collected using three checklists. In order to arrive at a consensus about the data elements, the decision Delphi technique was applied using questionnaires. The content validity and reliability of the questionnaires were assessed by experts' opinions and the test-retest method, respectively. An (MDS) of a traffic accident information management system was assigned to three sections: a minimum data set for traffic police with six classes, including 118 data elements; a trauma center with five data classes, including 57 data elements; and a medical emergency center, with 11 classes, including 64 data elements. Planning for the prevention of traffic accidents requires standardized data. As the foundation for crash prevention efforts, existing standard data infrastructures present policymakers and government officials with a great opportunity to strengthen and integrate existing accident information systems to better track road traffic injuries and fatalities.

  14. Persistence of airline accidents.

    Science.gov (United States)

    Barros, Carlos Pestana; Faria, Joao Ricardo; Gil-Alana, Luis Alberiko

    2010-10-01

    This paper expands on air travel accident research by examining the relationship between air travel accidents and airline traffic or volume in the period from 1927-2006. The theoretical model is based on a representative airline company that aims to maximise its profits, and it utilises a fractional integration approach in order to determine whether there is a persistent pattern over time with respect to air accidents and air traffic. Furthermore, the paper analyses how airline accidents are related to traffic using a fractional cointegration approach. It finds that airline accidents are persistent and that a (non-stationary) fractional cointegration relationship exists between total airline accidents and airline passengers, airline miles and airline revenues, with shocks that affect the long-run equilibrium disappearing in the very long term. Moreover, this relation is negative, which might be due to the fact that air travel is becoming safer and there is greater competition in the airline industry. Policy implications are derived for countering accident events, based on competition and regulation. © 2010 The Author(s). Journal compilation © Overseas Development Institute, 2010.

  15. Comparative modeling analyses of Cs-137 fate in the rivers impacted by Chernobyl and Fukushima accidents

    Energy Technology Data Exchange (ETDEWEB)

    Zheleznyak, M.; Kivva, S. [Institute of Environmental Radioactivity, Fukushima University (Japan)

    2014-07-01

    The consequences of two largest nuclear accidents of the last decades - at Chernobyl Nuclear Power Plant (ChNPP) (1986) and at Fukushima Daiichi NPP (FDNPP) (2011) clearly demonstrated that radioactive contamination of water bodies in vicinity of NPP and on the waterways from it, e.g., river- reservoir water after Chernobyl accident and rivers and coastal marine waters after Fukushima accident, in the both cases have been one of the main sources of the public concerns on the accident consequences. The higher weight of water contamination in public perception of the accidents consequences in comparison with the real fraction of doses via aquatic pathways in comparison with other dose components is a specificity of public perception of environmental contamination. This psychological phenomenon that was confirmed after these accidents provides supplementary arguments that the reliable simulation and prediction of the radionuclide dynamics in water and sediments is important part of the post-accidental radioecological research. The purpose of the research is to use the experience of the modeling activities f conducted for the past more than 25 years within the Chernobyl affected Pripyat River and Dnieper River watershed as also data of the new monitoring studies in Japan of Abukuma River (largest in the region - the watershed area is 5400 km{sup 2}), Kuchibuto River, Uta River, Niita River, Natsui River, Same River, as also of the studies on the specific of the 'water-sediment' {sup 137}Cs exchanges in this area to refine the 1-D model RIVTOX and 2-D model COASTOX for the increasing of the predictive power of the modeling technologies. The results of the modeling studies are applied for more accurate prediction of water/sediment radionuclide contamination of rivers and reservoirs in the Fukushima Prefecture and for the comparative analyses of the efficiency of the of the post -accidental measures to diminish the contamination of the water bodies. Document

  16. Development and application of a radioactivity evaluation technique the to obtain radiation exposure dose of radioactivity evaluation technique when a severe accident occurs in the a power station of a severe accident. Accident management guidelines of knowledge-based maintenance

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Yoshida, Yoshitaka

    2013-01-01

    As a One of the lessons learned from the nuclear accident at the Fukushima Daiichi Nuclear Power Stations of Tokyo Electric Power Company, the was the need for improvement of accident management guidelines is required. In this report study, we developed and applied a dose evaluation technique to evaluated the radiation dose in a nuclear power plant assuming three conditions: employees were evacuation evacuated at the time of a severe accident occurrence; operators carried out the accident management operation; of the operators, and the repair work was carried out for of the trouble damaged apparatuses in a the nuclear power plant using a dose evaluation system. The following knowledge findings were obtained and should to be reflected to in the knowledge base of the guidelines was obtained. (1) By making clearly identifying an areas beforehand becoming the that would receive high radiation doses at the time of a severe accident definitely beforehand, we can employees can be moved to the evacuation places through an areas having of low dose rate and it is also known it how much we long employees can safely stay in the evacuation places. (2) When they circulate CV containment vessel recirculation sump water is recirculated by for the accident management operation and the restoration of safety in the facilities, because the plumbing piping and the apparatuses become radioactive radioactivity sources, the dose evaluation of the shortest access route and detour access routes with should be made for effective the accident management operation is effective. Because the area where a dose rate rises changes which as safety apparatuses are restored, in consideration of a plant state, it is necessary to judge the rightness or wrongness of the work continuation from the spot radioactive dose of the actual apparatus area, with based on precedence of the need to restore with precedence, and to choose a system to be used for accident management. (author)

  17. Upon the reconstruction of accidents triggered by tire explosion. Analytical model and case study

    Science.gov (United States)

    Gaiginschi, L.; Agape, I.; Talif, S.

    2017-10-01

    Accident Reconstruction is important in the general context of increasing road traffic safety. In the casuistry of traffic accidents, those caused by tire explosions are critical under the severity of consequences, because they are usually happening at high speeds. Consequently, the knowledge of the running speed of the vehicle involved at the time of the tire explosion is essential to elucidate the circumstances of the accident. The paper presents an analytical model for the kinematics of a vehicle which, after the explosion of one of its tires, begins to skid, overturns and rolls. The model consists of two concurent approaches built as applications of the momentum conservation and energy conservation principles, and allows determination of the initial speed of the vehicle involved, by running backwards the sequences of the road event. The authors also aimed to both validate the two distinct analytical approaches by calibrating the calculation algorithms on a case study

  18. Fuel relocation modeling in the SAS4A accident analysis code system

    International Nuclear Information System (INIS)

    Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

    1986-01-01

    The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

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

    International Nuclear Information System (INIS)

    Chvetsov, I.; Kouznetsov, I.; Volkov, A.

    2000-01-01

    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)

  20. [Model of Analysis and Prevention of Accidents - MAPA: tool for operational health surveillance].

    Science.gov (United States)

    de Almeida, Ildeberto Muniz; Vilela, Rodolfo Andrade de Gouveia; da Silva, Alessandro José Nunes; Beltran, Sandra Lorena

    2014-12-01

    The analysis of work-related accidents is important for accident surveillance and prevention. Current methods of analysis seek to overcome reductionist views that see these occurrences as simple events explained by operator error. The objective of this paper is to analyze the Model of Analysis and Prevention of Accidents (MAPA) and its use in monitoring interventions, duly highlighting aspects experienced in the use of the tool. The descriptive analytical method was used, introducing the steps of the model. To illustrate contributions and or difficulties, cases where the tool was used in the context of service were selected. MAPA integrates theoretical approaches that have already been tried in studies of accidents by providing useful conceptual support from the data collection stage until conclusion and intervention stages. Besides revealing weaknesses of the traditional approach, it helps identify organizational determinants, such as management failings, system design and safety management involved in the accident. The main challenges lie in the grasp of concepts by users, in exploring organizational aspects upstream in the chain of decisions or at higher levels of the hierarchy, as well as the intervention to change the determinants of these events.

  1. Socioeconomic consequences of nuclear reactor accidents

    International Nuclear Information System (INIS)

    Tawil, J.J.; Callaway, J.W.; Coles, B.L.; Cronin, F.J.; Currie, J.W.; Imhoff, K.L.; Lewis, P.M.; Nesse, R.J.; Strenge, D.L.

    1984-06-01

    This report identifies and characterizes the off-site socioeconomic consequences that would likely result from a severe radiological accident at a nuclear power plant. The types of impacts that are addressed include economic impacts, health impacts, social/psychological impacts and institutional impacts. These impacts are identified for each of several phases of a reactor accident - from the warning phase through the post-resettlement phase. The relative importance of the impact during each accident phase and the degree to which the impact can be predicted are indicated. The report also examines the methods that are currently used for assessing nuclear reactor accidents, including development of accident scenarios and the estimating of socioeconomic accident consequences with various models. Finally, a critical evaluation is made regarding the use of impact analyses in estimating the contribution of socioeconomic consequences to nuclear accident reactor accident risk. 116 references, 7 figures, 15 tables

  2. Remediation strategies after nuclear or radiological accidents: part 1 - database development

    International Nuclear Information System (INIS)

    Silva, Diogo N.G.; Wasserman, Maria Angelica V.; Rochedo, Elaine R.R.

    2009-01-01

    The selection of protective measures and of remediation strategies of areas after a nuclear or radiological accident needs to be based on previously established criteria, in way to minimize the public's emotional stress and the exposure to workers involved in cleanup operations due to the implementation of procedures that are not effective in reducing doses to the public. Thus this work intended to develop a database which allows supporting the decision-making process after these accidents, by describing the foreseen strategies according to the type of accident and the type of affected environment, in order to be used in a multi-criteria selective process. To achieve that, in this first stage, the database has been developed including the following aspects: type of environment (urban, rural or aquatic); their contamination removal efficiency, as function of the time elapsed since the contamination event; the type and the amount of waste generated in the application of the strategy; the expected doses to the work team and basic needs such as specific materials, equipment, training, IPE, among others. The protection measures are usually described in literature considering their activity removal efficiency of a certain surface or environment. In order to determine their efficiency in the reduction of doses, a second stage is foreseen, involving the simulation of the implementation of the measures in different moments after the contamination, based on pre-defined accidents and scenarios, with focus on the surroundings of the Brazilian Nuclear Power Plants in Angra dos Reis. (author)

  3. Remediation strategies after nuclear or radiological accidents: part 1 - database development

    Energy Technology Data Exchange (ETDEWEB)

    Silva, Diogo N.G.; Wasserman, Maria Angelica V. [Instituto de Radioprotecao e Dosimetria (IRD/CNEN-RJ), Rio de Janeiro, RJ (Brazil)], e-mail: dneves@ird.gov.br, e-mail: angelica@ird.gov.br, e-mail: lfcconti@ird.gov.br; Rochedo, Elaine R.R. [Comissao Nacional de Energia Nuclear (CNEN-RJ), Rio de Janeiro, RJ (Brazil). Coordenacao de Instalacoes Nucleares], e-mail: erochedo@cnen.gov.br

    2009-07-01

    The selection of protective measures and of remediation strategies of areas after a nuclear or radiological accident needs to be based on previously established criteria, in way to minimize the public's emotional stress and the exposure to workers involved in cleanup operations due to the implementation of procedures that are not effective in reducing doses to the public. Thus this work intended to develop a database which allows supporting the decision-making process after these accidents, by describing the foreseen strategies according to the type of accident and the type of affected environment, in order to be used in a multi-criteria selective process. To achieve that, in this first stage, the database has been developed including the following aspects: type of environment (urban, rural or aquatic); their contamination removal efficiency, as function of the time elapsed since the contamination event; the type and the amount of waste generated in the application of the strategy; the expected doses to the work team and basic needs such as specific materials, equipment, training, IPE, among others. The protection measures are usually described in literature considering their activity removal efficiency of a certain surface or environment. In order to determine their efficiency in the reduction of doses, a second stage is foreseen, involving the simulation of the implementation of the measures in different moments after the contamination, based on pre-defined accidents and scenarios, with focus on the surroundings of the Brazilian Nuclear Power Plants in Angra dos Reis. (author)

  4. Assessment and prediction of road accident injuries trend using time-series models in Kurdistan.

    Science.gov (United States)

    Parvareh, Maryam; Karimi, Asrin; Rezaei, Satar; Woldemichael, Abraha; Nili, Sairan; Nouri, Bijan; Nasab, Nader Esmail

    2018-01-01

    Road traffic accidents are commonly encountered incidents that can cause high-intensity injuries to the victims and have direct impacts on the members of the society. Iran has one of the highest incident rates of road traffic accidents. The objective of this study was to model the patterns of road traffic accidents leading to injury in Kurdistan province, Iran. A time-series analysis was conducted to characterize and predict the frequency of road traffic accidents that lead to injury in Kurdistan province. The injuries were categorized into three separate groups which were related to the car occupants, motorcyclists and pedestrian road traffic accident injuries. The Box-Jenkins time-series analysis was used to model the injury observations applying autoregressive integrated moving average (ARIMA) and seasonal autoregressive integrated moving average (SARIMA) from March 2009 to February 2015 and to predict the accidents up to 24 months later (February 2017). The analysis was carried out using R-3.4.2 statistical software package. A total of 5199 pedestrians, 9015 motorcyclists, and 28,906 car occupants' accidents were observed. The mean (SD) number of car occupant, motorcyclist and pedestrian accident injuries observed were 401.01 (SD 32.78), 123.70 (SD 30.18) and 71.19 (SD 17.92) per year, respectively. The best models for the pattern of car occupant, motorcyclist, and pedestrian injuries were the ARIMA (1, 0, 0), SARIMA (1, 0, 2) (1, 0, 0) 12 , and SARIMA (1, 1, 1) (0, 0, 1) 12 , respectively. The motorcyclist and pedestrian injuries showed a seasonal pattern and the peak was during summer (August). The minimum frequency for the motorcyclist and pedestrian injuries were observed during the late autumn and early winter (December and January). Our findings revealed that the observed motorcyclist and pedestrian injuries had a seasonal pattern that was explained by air temperature changes overtime. These findings call the need for close monitoring of the

  5. Development of advanced claddings for suppressing the hydrogen emission in accident conditions. Development of advanced claddings for suppressing the hydrogen emission in the accident condition

    International Nuclear Information System (INIS)

    Park, Jeong-Yong; KIM, Hyun-Gil; JUNG, Yang-Il; PARK, Dong-Jun; KOO, Yang-Hyun

    2013-01-01

    The development of accident-tolerant fuels can be a breakthrough to help solve the challenge facing nuclear fuels. One of the goals to be reached with accident-tolerant fuels is to reduce the hydrogen emission in the accident condition by improving the high-temperature oxidation resistance of claddings. KAERI launched a new project to develop the accident-tolerant fuel claddings with the primary objective to suppress the hydrogen emission even in severe accident conditions. Two concepts are now being considered as hydrogen-suppressed cladding. In concept 1, the surface modification technique was used to improve the oxidation resistance of Zr claddings. Like in concept 2, the metal-ceramic hybrid cladding which has a ceramic composite layer between the Zr inner layer and the outer surface coating is being developed. The high-temperature steam oxidation behaviour was investigated for several candidate materials for the surface modification of Zr claddings. From the oxidation tests carried out in 1 200 deg. C steam, it was found that the high-temperature steam oxidation resistance of Cr and Si was much higher than that of zircaloy-4. Al 3 Ti-based alloys also showed extremely low-oxidation rate compared to zircaloy-4. One important part in the surface modification is to develop the surface coating technology where the optimum process needs to be established depending on the surface layer materials. Several candidate materials were coated on the Zr alloy specimens by a laser beam scanning (LBS), a plasma spray (PS) and a PS followed by LBS and subject to the high-temperature steam oxidation test. It was found that Cr and Si coating layers were effective in protecting Zr-alloys from the oxidation. The corrosion behaviour of the candidate materials in normal reactor operation condition such as 360 deg. C water will be investigated after the screening test in the high-temperature steam. The metal-ceramic hybrid cladding consisted of three major parts; a Zr liner, a

  6. Modeling the early-phase redistribution of radiocesium fallouts in an evergreen coniferous forest after Chernobyl and Fukushima accidents

    Energy Technology Data Exchange (ETDEWEB)

    Calmon, P.; Gonze, M.-A.; Mourlon, Ch.

    2015-10-01

    Following the Chernobyl accident, the scientific community gained numerous data on the transfer of radiocesium in European forest ecosystems, including information regarding the short-term redistribution of atmospheric fallout onto forest canopies. In the course of international programs, the French Institute for Radiological Protection and Nuclear Safety (IRSN) developed a forest model, named TREE4 (Transfer of Radionuclides and External Exposure in FORest systems), 15 years ago. Recently published papers on a Japanese evergreen coniferous forest contaminated by Fukushima radiocesium fallout provide interesting and quantitative data on radioactive mass fluxes measured within the forest in the months following the accident. The present study determined whether the approach adopted in the TREE4 model provides satisfactory results for Japanese forests or whether it requires adjustments. This study focused on the interception of airborne radiocesium by forest canopy, and the subsequent transfer to the forest floor through processes such as litterfall, throughfall, and stemflow, in the months following the accident. We demonstrated that TREE4 quite satisfactorily predicted the interception fraction (20%) and the canopy-to-soil transfer (70% of the total deposit in 5 months) in the Tochigi forest. This dynamics was similar to that observed in the Höglwald spruce forest. However, the unexpectedly high contribution of litterfall (31% in 5 months) in the Tochigi forest could not be reproduced in our simulations (2.5%). Possible reasons for this discrepancy are discussed; and sensitivity of the results to uncertainty in deposition conditions was analyzed. - Highlights: • Transfer of radiocesium atmospheric fallout in evergreen forests was modeled. • The model was tested using observations from Chernobyl and Fukushima accidents. • Model predictions of canopy interception and depuration agree with measurements. • Unexpectedly high contribution of litterfall for the

  7. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (5) Identification of dominant factors in ex-vessel accident sequences

    International Nuclear Information System (INIS)

    Ohno, Shuji; Seino, Hiroshi; Miyahara, Shinya

    2009-01-01

    The evaluation of accident progression outside of a reactor vessel (ex-vessel) and subsequent transfer behavior of radioactive materials is of great importance from the viewpoint of Level 2 PSA. Hence typical ex-vessel accident sequences in the JAEA Sodium-cooled Fast Reactor are qualitatively discussed in this paper and dominant behaviors or factors in the sequences are investigated through parametric calculations using the CONTAIN/LMR code. Scenarios to be focused on are, 1) sodium vapor leakage from the reactor vessel and 2) sodium-concrete reaction, which are both to be considered in the accident category of LOHRS (loss of heat removal system) and might be followed by an early containment failure due to the thermal effect of sodium combustion and hydrogen burning respectively. The calculated results clarify that the sodium vapor leak rate and the scale of sodium-concrete reaction are the important factors to dominate the ex-vessel accident progression. In addition to the understandings of the dominant factors, the analyzed results also provide the specific information such as pressure loading value to the containment and the timing of pressurization, which is indispensable as technical base in Level 2 PSA for developing event trees and for quantifying the accident consequences. (author)

  8. WASTE-ACC: A computer model for analysis of waste management accidents

    International Nuclear Information System (INIS)

    Nabelssi, B.K.; Folga, S.; Kohout, E.J.; Mueller, C.J.; Roglans-Ribas, J.

    1996-12-01

    In support of the U.S. Department of Energy's (DOE's) Waste Management Programmatic Environmental Impact Statement, Argonne National Laboratory has developed WASTE-ACC, a computational framework and integrated PC-based database system, to assess atmospheric releases from facility accidents. WASTE-ACC facilitates the many calculations for the accident analyses necessitated by the numerous combinations of waste types, waste management process technologies, facility locations, and site consolidation strategies in the waste management alternatives across the DOE complex. WASTE-ACC is a comprehensive tool that can effectively test future DOE waste management alternatives and assumptions. The computational framework can access several relational databases to calculate atmospheric releases. The databases contain throughput volumes, waste profiles, treatment process parameters, and accident data such as frequencies of initiators, conditional probabilities of subsequent events, and source term release parameters of the various waste forms under accident stresses. This report describes the computational framework and supporting databases used to conduct accident analyses and to develop source terms to assess potential health impacts that may affect on-site workers and off-site members of the public under various DOE waste management alternatives

  9. Part-time farmers and accidents with agricultural machinery: a moderated mediated model on the role played by frequency of use and unsafe beliefs.

    Science.gov (United States)

    Caffaro, Federica; Roccato, Michele; Micheletti Cremasco, Margherita; Cavallo, Eugenio

    2018-01-25

    We aimed at testing a model of the direct and indirect effects of being a part-time farmer on the probability of being involved in an agricultural machinery-related accident, considering the role played by unsafe beliefs and the frequency of use of machinery. Two-hundred and fifty-two Italian men, regular users of agricultural machinery (age: Mean = 45.1 years, standard Deviation = 17.5), were administered a paper-and-pencil questionnaire addressing their relation with work, unsafe beliefs, and previous experience of machinery-related accidents. Being a part-time farmer showed a positive association with unsafe beliefs only among occasional machinery users. Unsafe beliefs in turn showed a positive association with accidents. The study gave a novel contribution to the knowledge of the chain of events connecting part-time farmers with machinery-related accidents. Preventive training interventions targeting part-timer farmers using agricultural machinery just occasionally should be developed.

  10. Raim – A model for iodine behavior in containment under severe accident condition

    Directory of Open Access Journals (Sweden)

    Han-Chul Kim

    2015-12-01

    Full Text Available Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR, Organization for Economic Co-operation and Development (OECD-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model, based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  11. RAIM-A model for iodine behavior in containment under severe accident condition

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Han Chul; Cho, Yeong Hun [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2015-12-15

    Following a severe accident in a nuclear power plant, iodine is a major contributor to the potential health risks for the public. Because the amount of iodine released largely depends on its volatility, iodine's behavior in containment has been extensively studied in international programs such as International Source Term Programme-Experimental Program on Iodine Chemistry under Radiation (EPICUR), Organization for Economic Co-operation and Development (OECD)-Behaviour of Iodine Project, and OECD-Source Term Evaluation and Mitigation. Korea Institute of Nuclear Safety (KINS) has joined these programs and is developing a simplified, stand-alone iodine chemistry model, RAIM (Radio-Active Iodine chemistry Model), based on the IMOD methodology and other previous studies. This model deals with chemical reactions associated with the formation and destruction of iodine species and surface reactions in the containment atmosphere and the sump in a simple manner. RAIM was applied to a simulation of four EPICUR tests and one Radioiodine Test Facility test, which were carried out in aqueous or gaseous phases. After analysis, the results show a trend of underestimation of organic and molecular iodine for the gas-phase experiments, the opposite of that for the aqueous-phase ones, whereas the total amount of volatile iodine species agrees well between the experiment and the analysis result.

  12. [Prevention of occupational accidents with biological material as per Green and Kreuter Model].

    Science.gov (United States)

    Manetti, Marcela Luisa; da Costa, João Carlos Souza; Marziale, Maria Helena Palucci; Trovó, Marli Elisa

    2006-03-01

    This study aimed at diagnosing the occurrence of occupational accidents deriving from exposition to biological substance among workers of a hospital from São Paulo, Brazil, analyzing the adopted safety measures and elaborating a flowchart of preventive actions according to the Health Promotion Model by Green and Kreuter. It is an exploratory study with data collected electronically from the website REPAT - Electronic Network for the Prevention of Occupational Accidents with biological substances. The strategy used by the hospital did not reduce the injures. Results were used to elaborate a flowchart of preventive actions in order to improve the workers' quality of life.

  13. Maximum forseeable accident analysis made by a sodium leak on the BN-800 primary circuit and the more constraining accident development scenario

    International Nuclear Information System (INIS)

    Ivanenko, V.N.; Zybin, V.A.

    1988-01-01

    In this paper the different ways of development for the BN-800 maximum credible accident in case of loss and fire of primary sodium are examined. The more constraining scenario is presented. During the scenario analysis the accidental release of radioactive materials in the environment has been studied. These releases are below the authorized values [fr

  14. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  15. Impact of rainstorm and runoff modeling on predicted consequences of atmospheric releases from nuclear reactor accidents

    International Nuclear Information System (INIS)

    Ritchie, L.T.; Brown, W.D.; Wayland, J.R.

    1980-05-01

    A general temperate latitude cyclonic rainstorm model is presented which describes the effects of washout and runoff on consequences of atmospheric releases of radioactive material from potential nuclear reactor accidents. The model treats the temporal and spatial variability of precipitation processes. Predicted air and ground concentrations of radioactive material and resultant health consequences for the new model are compared to those of the original WASH-1400 model under invariant meteorological conditions and for realistic weather events using observed meteorological sequences. For a specific accident under a particular set of meteorological conditions, the new model can give significantly different results from those predicted by the WASH-1400 model, but the aggregate consequences produced for a large number of meteorological conditions are similar

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

    DEFF Research Database (Denmark)

    Kaplan, Sigal; Prato, Carlo Giacomo

    2012-01-01

    of 2011. Method: The current study investigates the underlying risk factors of bus accident severity in the United States by estimating a generalized ordered logit model. Data for the analysis are retrieved from the General Estimates System (GES) database for the years 2005–2009. Results: Results show...... that accident severity increases: (i) for young bus drivers under the age of 25; (ii) for drivers beyond the age of 55, and most prominently for drivers over 65 years old; (iii) for female drivers; (iv) for very high (over 65 mph) and very low (under 20 mph) speed limits; (v) at intersections; (vi) because......Introduction: Recent years have witnessed a growing interest in improving bus safety operations worldwide. While in the United States buses are considered relatively safe, the number of bus accidents is far from being negligible, triggering the introduction of the Motor-coach Enhanced Safety Act...

  17. The importance of trajectory modelling in accident consequence assessments

    International Nuclear Information System (INIS)

    Jones, J.A.; Williams, J.A.; Hill, M.D.

    1988-01-01

    Most atmospheric dispersion models used at present or probabilistic risk assessment (PRA) are linear: they take account of the wind speed but not the direction after the first hour. Therefore, the trajectory model is a more realistic description of the cloud's behaviour. However, the extra complexity means that the computing costs increase. This is an important factor for the MARIA code which is intended to be run on computers of varying power. The numbers of early effects predicted by a linear model and a trajectory model in a probabilistic risk assessment were compared to see which model should be preferred. The trajectory model predicted about 25% fewer expected early deaths and 30% more people evacuated than the linear model. However, the trajectory model took about ten times longer to calculate its results. The choice between the two models may depend on the speed of the computer available

  18. Aspects of uncertainty analysis in accident consequence modeling

    International Nuclear Information System (INIS)

    Travis, C.C.; Hoffman, F.O.

    1981-01-01

    Mathematical models are frequently used to determine probable dose to man from an accidental release of radionuclides by a nuclear facility. With increased emphasis on the accuracy of these models, the incorporation of uncertainty analysis has become one of the most crucial and sensitive components in evaluating the significance of model predictions. In the present paper, we address three aspects of uncertainty in models used to assess the radiological impact to humans: uncertainties resulting from the natural variability in human biological parameters; the propagation of parameter variability by mathematical models; and comparison of model predictions to observational data

  19. Development of the post-accident expert assessment capacity of I.R.S.N

    International Nuclear Information System (INIS)

    Repussard, J.

    2008-01-01

    Emergency preparedness is a key feature of IRSN's mission within the French institutional organisation of nuclear safety and radiation protection. In the event of an accident, IRSN is expected to propose to authorities the appropriate technical decisions in order to safeguard public health, environmental protection, and to restore safety. The article discusses the nature of the challenges that IRSN would have to meet in order to implement its mission during the 'post-accident' phase, which would last a long time if a significative environmental contamination did occur. These four challenges shape the strategy chosen by IRSN to prepare itself for such an eventuality: - to ensure the readiness of its experts, and the validity of its doctrine for intervention, on the basis of research results and of accumulated practical experience, - to maintain state of the art metrology facilities and computer models which would be needed to operate IRSN's response, - to elaborate, and maintain through exercises, plans to mobilize efficiently for the purpose of post-accident operations IRSN's resources which are normally dedicated to other tasks, - the work conducted by IRSN would also be expected to contribute to confidence rebuilding across society, after a nuclear accident. Communication and transparency will play a major role for this, and IRSN's culture and values have to take this fully into account. The conclusion stresses the importance of resource allocation across the different missions of IRSN, the achievement of prevention of accidents through the safety analysis and research processes remaining of paramount importance, and emergency preparedness being the last line of defence. International cooperation is in this respect one of the better ways to enhanced efficiency. (author)

  20. Study of heat and mass transfer phenomena in fuel assembly models under accident conditions

    International Nuclear Information System (INIS)

    Yefanov, A.D.; Kalyakin, C.G.; Loshchinin, V.M.; Pomet'ko, R.S.; Sergeev, V.V.; Shumsky, R.V.

    1996-01-01

    The majority of the material in support of the thermal - hydraulic safety of WWER core was obtained on single - assembly models containing a relatively small number of elements - heater rods. Upgrading the requirements to the reactor safety leads to the necessity for studying phenomena in channels representing the cross - sectional core dimensions and non - uniform radial power generation. Under such conditions, the contribution of natural convection can be significant in some core zones, including the occurrence of reverse flows and interchannel instability. These phenomena can have an important influence on heat transfer processes. Such influence is especially drastical under accident conditions associated with ceasing the forced circulation over the circuit. A number of urgent reactor safety problems at low operating parameters is related with the computer code verification and certification. One of the important trends in the reactor safety research is concerned with the rod bundle reflooding and verificational calculations of this phenomenon. To assess the water cooled reactor safety, the best fit computer codes are employed, which make it possible to simulate accident and transient operating conditions in a reactor installation. One of the most widely known computer codes is the RELAP5/MOD3 Code. The paper presents the comparison of the results calculated using this computer code with the test data on 4 - rod bundle quenching, which were obtained at the SSCRF-IPPE. Recently, the investigations on the steam - zirconium reaction kinetics have been performed at the SSCFR-IPPE and are being presently performed for the purpose of developing new and verifying available computer codes. (author). 3 refs, 6 figs

  1. The study on development of emergency operating procedures based on symptom and risk for accident management

    International Nuclear Information System (INIS)

    Kang, K. S.; Jeong, H. J.

    1998-01-01

    The Advanced EOP(AEOP) has been developed by focusing on the importance of the operators role in emergency conditions. In the AEOP, to overcome the complexity of current EOPs and maintain the consistency of operators action according to plant emergency conditions, operator's task were allocated according to their duties. As an alternative, the Computerized Operator Aid System (COAS) has been developed to reduce operator's burden and provide detailed instructions of procedure. Probabilistic Safety Assessment (PSA) results were synthesized in the AEOP using the event tree to give the awareness and the prediction of accident progression in advance. In conclusion, the existing EOP with its inherent complexity should be simplified and consolidated using computerized operator support system and task allocation to prevent more severe accidents and to reduce operator cognitive overload in emergency conditions

  2. The study on development of emergency operating procedures based on symptom and risk for accident management

    Energy Technology Data Exchange (ETDEWEB)

    Kang, K. S.; Jeong, H. J. [KOPEC, Taejon (Korea, Republic of)

    1998-10-01

    The Advanced EOP(AEOP) has been developed by focusing on the importance of the operators role in emergency conditions. In the AEOP, to overcome the complexity of current EOPs and maintain the consistency of operators action according to plant emergency conditions, operator's task were allocated according to their duties. As an alternative, the Computerized Operator Aid System (COAS) has been developed to reduce operator's burden and provide detailed instructions of procedure. Probabilistic Safety Assessment (PSA) results were synthesized in the AEOP using the event tree to give the awareness and the prediction of accident progression in advance. In conclusion, the existing EOP with its inherent complexity should be simplified and consolidated using computerized operator support system and task allocation to prevent more severe accidents and to reduce operator cognitive overload in emergency conditions.

  3. Development and application of the practice tool to deal with severe accident

    International Nuclear Information System (INIS)

    Kawasaki, Ikuo; Yoshida, Yoshitaka; Iwasaki, Yoshito

    2014-01-01

    We developed the practice tool to simulate communications between operators at a nuclear power station and persons at the headquarters at the time of severe accident (SA). The tool was developed from considering the lessons learned in dealing with the accident at the Tokyo Electric Power Company Fukushima Daiichi Nuclear Power Station, especially related to making appropriate responses to events. The tool allows users at headquarters to learn about the constitution of a specific plant and to make a reply and state a judgment based on knowledge about SA. The situations used for the practice tool were made using SPDS data from past disaster prevention drills. In a test, SA education of headquarters workers was carried out using this practice tool, and we confirmed that users were able to make the right phenomenon judgment and communicate it effectively based on the knowledge given by this practice tool. (author)

  4. Analyses of non-fatal accidents in an opencast mine by logistic regression model - a case study.

    Science.gov (United States)

    Onder, Seyhan; Mutlu, Mert

    2017-09-01

    Accidents cause major damage for both workers and enterprises in the mining industry. To reduce the number of occupational accidents, these incidents should be properly registered and carefully analysed. This study efficiently examines the Aegean Lignite Enterprise (ELI) of Turkish Coal Enterprises (TKI) in Soma between 2006 and 2011, and opencast coal mine occupational accident records were used for statistical analyses. A total of 231 occupational accidents were analysed for this study. The accident records were categorized into seven groups: area, reason, occupation, part of body, age, shift hour and lost days. The SPSS package program was used in this study for logistic regression analyses, which predicted the probability of accidents resulting in greater or less than 3 lost workdays for non-fatal injuries. Social facilities-area of surface installations, workshops and opencast mining areas are the areas with the highest probability for accidents with greater than 3 lost workdays for non-fatal injuries, while the reasons with the highest probability for these types of accidents are transporting and manual handling. Additionally, the model was tested for such reported accidents that occurred in 2012 for the ELI in Soma and estimated the probability of exposure to accidents with lost workdays correctly by 70%.

  5. Accident sequence quantification with KIRAP

    International Nuclear Information System (INIS)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

  6. Accident sequence quantification with KIRAP

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

    1997-01-01

    The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

  7. Concepts in developing technical means of accident shutdown of nuclear reactor

    International Nuclear Information System (INIS)

    Ionajtis, R.R.; Mikhajlov, M.P.; Cherkashov, Yu.M.

    1992-01-01

    Logic for realization of multistage (echelon) reactor accident shutdown system (ASS) is proposed on the basis of general safety concepts (OPB-88). ASS includes the basis stage with traditional composition of member systems (executive, control, providing ones), auxiliary (doubling) on the other principle of action and insuring (with direct action). Structural schemes of the system as a whole and member subsystems are presented. Recommendations on developing executive and control subsystems are given

  8. DEVELOPMENT OF METHODOLOGY FOR TRAFFIC ACCIDENT FORECASTING AT VARIOUS TYPICAL URBAN AREAS

    OpenAIRE

    D. V. Kapsky

    2012-01-01

    The paper provides investigation results pertaining to development of methodology for forecasting traffic accidents using a “conflict zone” method that considers potential danger for two typical urban areas, namely: signaled crossings and bumps that are made in the areas of zebra crossings and it also considers various types and kinds of conflicts. The investigations have made it possible to obtain various indices of threshold sensitivity in respect of  potential risks  and in relation to tra...

  9. Development of criticality accident detector measuring neutrons and gamma-rays

    International Nuclear Information System (INIS)

    Tsujimura, Norio; Yoshida, Tadayoshi; Ishii, Masato

    2005-01-01

    The authors developed a new criticality accident detector measuring neutrons and gamma-rays. The detector is a cylindrical plastic scintillator coupled to a current-mode operated photomultiplier, and is covered by an inner cadmium shell, acting as a neutron to gamma-ray converter, and a 5cm thick outer polyethylene moderator in order to respond to the same threshold triggering dose regardless of whether it was exposed to neutrons, gamma-rays or a mixture of the two radiations. (author)

  10. The hierarchy-by-interval approach to identifying important models that need improvement in severe-accident simulation codes

    International Nuclear Information System (INIS)

    Heames, T.J.; Khatib-Rahbar, M.; Kelly, J.E.

    1995-01-01

    The hierarchy-by-interval (HBI) methodology was developed to determine an appropriate phenomena identification and ranking table for an independent peer review of severe-accident computer codes. The methodology is described, and the results of a specific code review are presented. Use of this systematic and structured approach ensures that important code models that need improvement are identified and prioritized, which allows code sponsors to more effectively direct limited resources in future code development. In addition, critical phenomenological areas that need more fundamental work, such as experimentation, are identified

  11. The usefulness of time-dependent reactor accident consequence modelling for emergency response planning

    International Nuclear Information System (INIS)

    Paretzke, H.G.; Jacob, P.; Mueller, H.; Proehl, G.

    1989-01-01

    After major releases of radionuclides into the atmosphere fast reaction of authorities will be necessary to inform the public of potential consequences and to consider and optimize mitigating actions. These activities require availability of well designed computer models, adequate and fast measurements and prior training of responsible persons. The quantitative assessment models should be capable of taking into account of actual atmospheric dispersion conditions, actual deposition situation (dry, rain, snow, fog), seasonal status of the agriculture, food processing and distribution pathways, etc. In this paper the usefulness of such models will be discussed, their limitations, the relative importance of exposure pathways and a selection of important methods to decrease the activity in food products after an accident. Real-time reactor accident consequence models should be considered as a condition sine qua non for responsible use of nuclear power for electricity production

  12. Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents. Part 4: IDAC causal model of operator problem-solving response

    Energy Technology Data Exchange (ETDEWEB)

    Chang, Y.H.J. [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States) and Paul Scherrer Institute, 5232 Villigen PSI (Switzerland)]. E-mail: yhc@umd.edu; Mosleh, A. [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States)

    2007-08-15

    This is the fourth in a series of five papers describing the Information, Decision, and Action in Crew context (IDAC) operator response model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model has been developed to probabilistically predicts the responses of a nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper assesses the effects of the performance-influencing factors (PIFs) affecting the operators' problem-solving responses including information pre-processing (I), diagnosis and decision making (D), and action execution (A). Literature support and justifications are provided for the assessment on the influences of PIFs.

  13. Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents. Part 4: IDAC causal model of operator problem-solving response

    International Nuclear Information System (INIS)

    Chang, Y.H.J.; Mosleh, A.

    2007-01-01

    This is the fourth in a series of five papers describing the Information, Decision, and Action in Crew context (IDAC) operator response model for human reliability analysis. An example application of this modeling technique is also discussed in this series. The model has been developed to probabilistically predicts the responses of a nuclear power plant control room operating crew in accident conditions. The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper assesses the effects of the performance-influencing factors (PIFs) affecting the operators' problem-solving responses including information pre-processing (I), diagnosis and decision making (D), and action execution (A). Literature support and justifications are provided for the assessment on the influences of PIFs

  14. Oil Spill Detection and Modelling: Preliminary Results for the Cercal Accident

    Science.gov (United States)

    da Costa, R. T.; Azevedo, A.; da Silva, J. C. B.; Oliveira, A.

    2013-03-01

    Oil spill research has significantly increased mainly as a result of the severe consequences experienced from industry accidents. Oil spill models are currently able to simulate the processes that determine the fate of oil slicks, playing an important role in disaster prevention, control and mitigation, generating valuable information for decision makers and the population in general. On the other hand, satellite Synthetic Aperture Radar (SAR) imagery has demonstrated significant potential in accidental oil spill detection, when they are accurately differentiated from look-alikes. The combination of both tools can lead to breakthroughs, particularly in the development of Early Warning Systems (EWS). This paper presents a hindcast simulation of the oil slick resulting from the Motor Tanker (MT) Cercal oil spill, listed by the Portuguese Navy as one of the major oil spills in the Portuguese Atlantic Coast. The accident took place nearby Leix˜oes Harbour, North of the Douro River, Porto (Portugal) on the 2nd of October 1994. The oil slick was segmented from available European Remote Sensing (ERS) satellite SAR images, using an algorithm based on a simplified version of the K-means clustering formulation. The image-acquired information, added to the initial conditions and forcings, provided the necessary inputs for the oil spill model. Simulations were made considering the tri-dimensional hydrodynamics in a crossscale domain, from the interior of the Douro River Estuary to the open-ocean on the Iberian Atlantic shelf. Atmospheric forcings (from ECMWF - the European Centre for Medium-Range Weather Forecasts and NOAA - the National Oceanic and Atmospheric Administration), river forcings (from SNIRH - the Portuguese National Information System of the Hydric Resources) and tidal forcings (from LNEC - the National Laboratory for Civil Engineering), including baroclinic gradients (NOAA), were considered. The lack of data for validation purposes only allowed the use of the

  15. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    International Nuclear Information System (INIS)

    Verfondern, K.; Mueller, D.

    1991-01-01

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  16. Improved point-kinetics model for the BWR control rod drop accident

    International Nuclear Information System (INIS)

    Neogy, P.; Wakabayashi, T.; Carew, J.F.

    1985-01-01

    A simple prescription to account for spatial feedback weighting effects in RDA (rod drop accident) point-kinetics analyses has been derived and tested. The point-kinetics feedback model is linear in the core peaking factor, F/sub Q/, and in the core average void fraction and fuel temperature. Comparison with detailed spatial kinetics analyses indicates that the improved point-kinetics model provides an accurate description of the BWR RDA

  17. Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

    Energy Technology Data Exchange (ETDEWEB)

    Le Pallec, J. C.; Crouzet, N.; Bergeaud, V.; Delavaud, C. [CEA/DEN/DM2S, CEA/Saclay, 91191 Gif sur Yvette Cedex (France)

    2012-07-01

    The control of uncertainties in the field of reactor physics and their propagation in best-estimate modeling are a major issue in safety analysis. In this framework, the CEA develops a methodology to perform multi-physics simulations including uncertainties analysis. The present paper aims to present and apply this methodology for the analysis of an accidental situation such as REA (Rod Ejection Accident). This accident is characterized by a strong interaction between the different areas of the reactor physics (neutronic, fuel thermal and thermal hydraulic). The modeling is performed with CRONOS2 code. The uncertainties analysis has been conducted with the URANIE platform developed by the CEA: For each identified response from the modeling (output) and considering a set of key parameters with their uncertainties (input), a surrogate model in the form of a neural network has been produced. The set of neural networks is then used to carry out a sensitivity analysis which consists on a global variance analysis with the determination of the Sobol indices for all responses. The sensitivity indices are obtained for the input parameters by an approach based on the use of polynomial chaos. The present exercise helped to develop a methodological flow scheme, to consolidate the use of URANIE tool in the framework of parallel calculations. Finally, the use of polynomial chaos allowed computing high order sensitivity indices and thus highlighting and classifying the influence of identified uncertainties on each response of the analysis (single and interaction effects). (authors)

  18. Severe accident training simulator APROS SA

    International Nuclear Information System (INIS)

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

    2003-01-01

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

  19. Accident information needs

    International Nuclear Information System (INIS)

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

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

  20. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-12-31

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  1. Accident information needs

    Energy Technology Data Exchange (ETDEWEB)

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

    1992-01-01

    A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

  2. Modeling Loss-of-Flow Accidents and Their Impact on Radiation Heat Transfer

    Directory of Open Access Journals (Sweden)

    Jivan Khatry

    2017-01-01

    Full Text Available Long-term high payload missions necessitate the need for nuclear space propulsion. The National Aeronautics and Space Administration (NASA investigated several reactor designs from 1959 to 1973 in order to develop the Nuclear Engine for Rocket Vehicle Application (NERVA. Study of planned/unplanned transients on nuclear thermal rockets is important due to the need for long-term missions. In this work, a system model based on RELAP5 is developed to simulate loss-of-flow accidents on the Pewee I test reactor. This paper investigates the radiation heat transfer between the fuel elements and the structures around it. In addition, the impact on the core fuel element temperature and average core pressure was also investigated. The following expected results were achieved: (i greater than normal fuel element temperatures, (ii fuel element temperatures exceeding the uranium carbide melting point, and (iii average core pressure less than normal. Results show that the radiation heat transfer rate between fuel elements and cold surfaces increases with decreasing flow rate through the reactor system. However, radiation heat transfer decreases when there is a complete LOFA. When there is a complete LOFA, the peripheral coolant channels of the fuel elements handle most of the radiation heat transfer. A safety system needs to be designed to counteract the decay heat resulting from a post-LOFA reactor scram.

  3. Model to predict radiological consequences of transportation accidents involving dispersal of radioactive material in urban areas

    International Nuclear Information System (INIS)

    Taylor, J.M.; Daniel, S.L.

    1978-01-01

    The analysis of accidental releases of radioactive material which may result from transportation accidents in high-density urban areas is influenced by several urban characteristics which make computer simulation the calculational method of choice. These urban features fall into four categories. Each of these categories contains time- and location-dependent parameters which must be coupled to the actual time and location of the release in the calculation of the anticipated radiological consequences. Due to the large number of dependent parameters a computer model, METRAN, has been developed to quantify these radiological consequences. Rather than attempt to describe an urban area as a single entity, a specific urban area is subdivided into a set of cells of fixed size to permit more detailed characterization. Initially, the study area is subdivided into a set of 2-dimensional cells. A uniform set of time-dependent physical characteristics which describe the land use, population distribution, traffic density, etc., within that cell are then computed from various data sources. The METRAN code incorporates several details of urban areas. A principal limitation of the analysis is the limited availability of accurate information to use as input data. Although the code was originally developed to analyze dispersal of radioactive material, it is currently being evaluated for use in analyzing the effects of dispersal of other hazardous materials in both urban and rural areas

  4. Partial least square method for modelling ergonomic risks factors on express bus accidents in the east coast of peninsular west Malaysia

    Energy Technology Data Exchange (ETDEWEB)

    Hashim, Yusof bin [Faculty of Industrial Sciences and Technology, Universiti Malaysia Pahang, Gambang 26300 Kuantan, Pahang (Malaysia); Taha, Zahari bin [Faculty of Manufacturing Engineering, Malaysia Pahang, 26600 Pekan, Pahang (Malaysia)

    2015-02-03

    Public, stake holders and authorities in Malaysian government show great concern towards high numbers of passenger’s injuries and passengers fatalities in express bus accident. This paper studies the underlying factors involved in determining ergonomics risk factors towards human error as the reasons in express bus accidents in order to develop an integrated analytical framework. Reliable information about drivers towards bus accident should lead to the design of strategies intended to make the public feel safe in public transport services. In addition there is an analysis of ergonomics risk factors to determine highly ergonomic risk factors which led to accidents. The research was performed in east coast of peninsular Malaysia using variance-based structural equation modeling namely the Partial Least Squares (PLS) regression techniques. A questionnaire survey was carried out at random among 65 express bus drivers operating from the city of Kuantan in Pahang and among 49 express bus drivers operating from the city of Kuala Terengganu in Terengganu to all towns in the east coast of peninsular west Malaysia. The ergonomic risks factors questionnaire is based on demographic information, occupational information, organizational safety climate, ergonomic workplace, physiological factors, stress at workplace, physical fatigue and near miss accidents. The correlation and significant values between latent constructs (near miss accident) were analyzed using SEM SmartPLS, 3M. The finding shows that the correlated ergonomic risks factors (occupational information, t=2.04, stress at workplace, t = 2.81, physiological factor, t=2.08) are significant to physical fatigue and as the mediator to near miss accident at t = 2.14 at p<0.05and T-statistics, t>1.96. The results shows that the effects of physical fatigue due to ergonomic risks factors influence the human error as the reasons in express bus accidents.

  5. Partial least square method for modelling ergonomic risks factors on express bus accidents in the east coast of peninsular west Malaysia

    International Nuclear Information System (INIS)

    Hashim, Yusof bin; Taha, Zahari bin

    2015-01-01

    Public, stake holders and authorities in Malaysian government show great concern towards high numbers of passenger’s injuries and passengers fatalities in express bus accident. This paper studies the underlying factors involved in determining ergonomics risk factors towards human error as the reasons in express bus accidents in order to develop an integrated analytical framework. Reliable information about drivers towards bus accident should lead to the design of strategies intended to make the public feel safe in public transport services. In addition there is an analysis of ergonomics risk factors to determine highly ergonomic risk factors which led to accidents. The research was performed in east coast of peninsular Malaysia using variance-based structural equation modeling namely the Partial Least Squares (PLS) regression techniques. A questionnaire survey was carried out at random among 65 express bus drivers operating from the city of Kuantan in Pahang and among 49 express bus drivers operating from the city of Kuala Terengganu in Terengganu to all towns in the east coast of peninsular west Malaysia. The ergonomic risks factors questionnaire is based on demographic information, occupational information, organizational safety climate, ergonomic workplace, physiological factors, stress at workplace, physical fatigue and near miss accidents. The correlation and significant values between latent constructs (near miss accident) were analyzed using SEM SmartPLS, 3M. The finding shows that the correlated ergonomic risks factors (occupational information, t=2.04, stress at workplace, t = 2.81, physiological factor, t=2.08) are significant to physical fatigue and as the mediator to near miss accident at t = 2.14 at p<0.05and T-statistics, t>1.96. The results shows that the effects of physical fatigue due to ergonomic risks factors influence the human error as the reasons in express bus accidents

  6. Partial least square method for modelling ergonomic risks factors on express bus accidents in the east coast of peninsular west Malaysia

    Science.gov (United States)

    Hashim, Yusof bin; Taha, Zahari bin

    2015-02-01

    Public, stake holders and authorities in Malaysian government show great concern towards high numbers of passenger's injuries and passengers fatalities in express bus accident. This paper studies the underlying factors involved in determining ergonomics risk factors towards human error as the reasons in express bus accidents in order to develop an integrated analytical framework. Reliable information about drivers towards bus accident should lead to the design of strategies intended to make the public feel safe in public transport services. In addition there is an analysis of ergonomics risk factors to determine highly ergonomic risk factors which led to accidents. The research was performed in east coast of peninsular Malaysia using variance-based structural equation modeling namely the Partial Least Squares (PLS) regression techniques. A questionnaire survey was carried out at random among 65 express bus drivers operating from the city of Kuantan in Pahang and among 49 express bus drivers operating from the city of Kuala Terengganu in Terengganu to all towns in the east coast of peninsular west Malaysia. The ergonomic risks factors questionnaire is based on demographic information, occupational information, organizational safety climate, ergonomic workplace, physiological factors, stress at workplace, physical fatigue and near miss accidents. The correlation and significant values between latent constructs (near miss accident) were analyzed using SEM SmartPLS, 3M. The finding shows that the correlated ergonomic risks factors (occupational information, t=2.04, stress at workplace, t = 2.81, physiological factor, t=2.08) are significant to physical fatigue and as the mediator to near miss accident at t = 2.14 at p1.96. The results shows that the effects of physical fatigue due to ergonomic risks factors influence the human error as the reasons in express bus accidents.

  7. Development of filter module for passive filtration and accident gas release confinement system for NPP

    International Nuclear Information System (INIS)

    Yelizarov, P.G.; Efanov, A.D.; Martynov, P.N.; Masalov, D.P.; Osipov, V.P.; Yagodkin, I.V.

    2005-01-01

    Full text of publication follows: One of the urgent problems of the safe NPP operation is air cleaning from radioactive aerosols and volatile iodine compounds under the accident operation conditions of NPP. A principally new passive accident gas release confinement system is used as the basis of the designs of new generation reactor power blocks under the-beyond-design-basis accident conditions with total loss of current. The basic structural component of the passive filtration system (PFS) is the filter-sorber being heated up to 300 deg. C. The filter-sorber represents a design consisting of 150 connected in parallel two-step filtering modules. The first step is intended to clean air from radioactive aerosols, the second one - to clean air from radioactive iodine and its volatile compounds. The filter-sorber is located in the upper point of the exterior protection shell. Due to natural convection, it provides confinement of r/a impurities and controlled steam-gas release from the inter-shell space into atmosphere. The basic specific design feature is the two-section design of the PFS filter module consisting of a coarse-cleaning section and a fine-cleaning section. A combination of layer-by-layer put filtering materials on the basis of glass fiber and metal fiber. The pilot PFS filter module specimen tests run in conditions modeling accident situation indicated that at a filtration rate of 0,3 cm/s the aerodynamic resistance of the module does not exceed 12 Pa, the filtration effectiveness equals 99,99 % in terms of aerosol, no less than 99,9% in terms of radioactive 131 I and no less than 99,0% in terms of organic compounds of iodine (CH 3 131 I); the dust capacity amounts to a value above 50 g/m 2 . The obtained results of tests comply with the design requirements imposed on the PFS filter-sorber module. (authors)

  8. A comparative analysis of reactor lower head debris cooling models employed in the existing severe accident analysis codes

    International Nuclear Information System (INIS)

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

    1998-08-01

    MELCOR and MAAP4 are the representative severe accident analysis codes which have been developed for the integral analysis of the phenomenological reactor lower head corium cooling behavior. Main objectives of the present study is to identify merits and disadvantages of each relevant model through the comparative analysis of the lower plenum corium cooling models employed in these two codes. The final results will be utilized for the development of LILAC phenomenological models and for the continuous improvement of the existing MELCOR reactor lower head models, which are currently being performed at the KAERI. For these purposes, first, nine reference models are selected featuring the lower head corium behavior based on the existing experimental evidences and related models. Then main features of the selected models have been critically analyzed, and finally merits and disadvantages of each corresponding model have been summarized in the view point of realistic corium behavior and reasonable modeling. Being on these evidences, summarized and presented the potential improvements for developing more advanced models. The present study has been focused on the qualitative comparison of each model and so more detailed quantitative analysis is strongly required to obtain the final conclusions for their merits and disadvantages. In addition, in order to compensate the limitations of the current model, required further studies relating closely the detailed mechanistic models with the molten material movement and heat transfer based on phase-change in the porous medium, to the existing simple models. (author). 36 refs

  9. Characteristics of the development of the radiological situation resulting from the accident, intervention levels and countermeasures

    International Nuclear Information System (INIS)

    Belyaev, S.T.; Demin, V.F.; Kutkov, V.A.; Bariakhtar, V.G.; Petriaev, E.P.

    1996-01-01

    Great efforts have been made in the frame of the national and international research programs to get complete data on the radioactive releases, environmental contamination and radiological situation resulted from the Chernobyl accident. Beginning from the first publication (IAEA meeting, August 1986) these data have been considerably improved and added. The most important change of them with their influence on the decision making in the mitigation activity and the current situation is described and analyzed. The national and international regulatory documents at the moment of the accident were neither complete nor perfect in some necessary aspects especially in respect to the countermeasures at the intermediate and long-term phases. New documents have been worked out during the intervention activity. From 1986 series of documents were developed on the national and international levels. These documents are considered and analyzed in the context of their practical implementation and by the modern experience and research results. The history of countermeasures adopted on the different intervention phases are described. These documents mainly establish intervention levels in terms of averted doses and regulate only radiation protection. They don't content any intervention levels in terms of residual doses and risk, which are necessary for regulation of social and health protection of population suffered from the accident. Other restriction for the optimal regulation comes from use of the effective dose for establishing intervention levels. These and other respective aspects are discussed

  10. The development of an inherent safety approach to the prevention of domino accidents.

    Science.gov (United States)

    Cozzani, Valerio; Tugnoli, Alessandro; Salzano, Ernesto

    2009-11-01

    The severity of industrial accidents in which a domino effect takes place is well known in the chemical and process industry. The application of an inherent safety approach for the prevention of escalation events leading to domino accidents was explored in the present study. Reference primary scenarios were analyzed and escalation vectors were defined. Inherent safety distances were defined and proposed as a metric to express the intensity of the escalation vectors. Simple rules of thumb were presented for a preliminary screening of these distances. Swift reference indices for layout screening with respect to escalation hazard were also defined. Two case studies derived from existing layouts of oil refineries were selected to understand the potentialities coming from the application in the methodology. The results evidenced that the approach allows a first comparative assessment of the actual domino hazard in a layout, and the identification of critical primary units with respect to escalation events. The methodology developed also represents a useful screening tool to identify were to dedicate major efforts in the design of add-on measures, optimizing conventional passive and active measures for the prevention of severe domino accidents.

  11. Application of the Life Change Unit model for the prevention of accident proneness among small to medium sized industries in Korea.

    Science.gov (United States)

    Kang, Youngsig; Hahm, Hyojoon; Yang, Sunghwan; Kim, Taegu

    2008-10-01

    Behavior models have provided an accident proneness concept based on life change unit (LCU) factors. This paper describes the development of a Korean Life Change Unit (KLCU) model for workers and managers in fatal accident areas, as well as an evaluation of its application. Results suggest that death of parents is the highest stress-giving factor for employees of small and medium sized industries a rational finding the viewpoint of Korean culture. The next stress-giving factors were shown to be the death of a spouse or loved ones, followed by the death of close family members, the death of close friends, changes of family members' health, unemployment, and jail terms. It turned out that these factors have a serious effect on industrial accidents and work-related diseases. The death of parents and close friends are ranked higher in the KLCU model than that of Western society. Crucial information for industrial accident prevention in real fields will be provided and the provided information will be useful for safety management programs related to accident prevention.

  12. Identification of Loss-of-Coolant Accidents in LWRs by Inverse Models

    International Nuclear Information System (INIS)

    Cholewa, Wojciech; Frid, Wiktor; Bednarski, Marcin

    2004-01-01

    This paper describes a novel diagnostic method based on inverse models that could be applied to identification of transients and accidents in nuclear power plants. In particular, it is shown that such models could be successfully applied to identification of loss-of-coolant accidents (LOCAs). This is demonstrated for LOCA scenarios for a boiling water reactor. Two classes of inverse models are discussed: local models valid only in a selected neighborhood of an unknown element in the data set, representing a state of a considered object, and global models, in the form of partially unilateral models, valid over the whole learning data set. An interesting and useful property of local inverse models is that they can be considered as example-based models, i.e., models that are spanned on particular sets of pattern data. It is concluded that the optimal diagnostic method should combine the advantages of both models, i.e., the high quality of results obtained from a local inverse model and the information about the confidence interval for the expected output provided by a partially unilateral model

  13. Human reliability data, human error and accident models--illustration through the Three Mile Island accident analysis

    International Nuclear Information System (INIS)

    Le Bot, Pierre

    2004-01-01

    Our first objective is to provide a panorama of Human Reliability data used in EDF's Safety Probabilistic Studies, and then, since these concepts are at the heart of Human Reliability and its methods, to go over the notion of human error and the understanding of accidents. We are not sure today that it is actually possible to provide in this field a foolproof and productive theoretical framework. Consequently, the aim of this article is to suggest potential paths of action and to provide information on EDF's progress along those paths which enables us to produce the most potentially useful Human Reliability analyses while taking into account current knowledge in Human Sciences. The second part of this article illustrates our point of view as EDF researchers through the analysis of the most famous civil nuclear accident, the Three Mile Island unit accident in 1979. Analysis of this accident allowed us to validate our positions regarding the need to move, in the case of an accident, from the concept of human error to that of systemic failure in the operation of systems such as a nuclear power plant. These concepts rely heavily on the notion of distributed cognition and we will explain how we applied it. These concepts were implemented in the MERMOS Human Reliability Probabilistic Assessment methods used in the latest EDF Probabilistic Human Reliability Assessment. Besides the fact that it is not very productive to focus exclusively on individual psychological error, the design of the MERMOS method and its implementation have confirmed two things: the significance of qualitative data collection for Human Reliability, and the central role held by Human Reliability experts in building knowledge about emergency operation, which in effect consists of Human Reliability data collection. The latest conclusion derived from the implementation of MERMOS is that, considering the difficulty in building 'generic' Human Reliability data in the field we are involved in, the best

  14. Development of Electrical Capacitance Sensors for Accident Tolerant Fuel (ATF) Testing at the Transient Reactor Test (TREAT) Facility

    Energy Technology Data Exchange (ETDEWEB)

    Liu, Maolong; Ryals, Matthew; Ali, Amir; Blandford, Edward; Jensen, Colby; Condie, Keith; Svoboda, John; O' Brien, Robert

    2016-08-01

    A variety of instruments are being developed and qualified to support the Accident Tolerant Fuels (ATF) program and future transient irradiations at the Transient Reactor Test (TREAT) facility at Idaho National Laboratory (INL). The University of New Mexico (UNM) is working with INL to develop capacitance-based void sensors for determining the timing of critical boiling phenomena in static capsule fuel testing and the volume-averaged void fraction in flow-boiling in-pile water loop fuel testing. The static capsule sensor developed at INL is a plate-type configuration, while UNM is utilizing a ring-type capacitance sensor. Each sensor design has been theoretically and experimentally investigated at INL and UNM. Experiments are being performed at INL in an autoclave to investigate the performance of these sensors under representative Pressurized Water Reactor (PWR) conditions in a static capsule. Experiments have been performed at UNM using air-water two-phase flow to determine the sensitivity and time response of the capacitance sensor under a flow boiling configuration. Initial measurements from the capacitance sensor have demonstrated the validity of the concept to enable real-time measurement of void fraction. The next steps include designing the cabling interface with the flow loop at UNM for Reactivity Initiated Accident (RIA) ATF testing at TREAT and further characterization of the measurement response for each sensor under varying conditions by experiments and modeling.

  15. Improved atmospheric dispersion modelling in the new program system UFOMOD for accident consequence assessments

    International Nuclear Information System (INIS)

    Panitz, H.J.

    1988-01-01

    An essential aim of the improvements of the new program system UFOMOD for Accident Consequence Assessments (ACAs) was to substitute the straightline Gaussian plume model conventionally used in ACA models by more realistic atmospheric dispersion models. To identify improved models which can be applied in ACA codes and to quantify the implications of different concepts of dispersion modelling on the results of an ACA, probabilistic comparative calculations with different atmospheric dispersion models have been carried out. The study showed that there are trajectory models available which can be applied in ACAs and that these trajectory models provide more realistic results of ACAs than straight-line Gaussian models. This led to a completly novel concept of atmospheric dispersion modelling which distinguish between two different distance ranges of validity: the near range ( 50 km). The two ranges are assigned to respective trajectory models

  16. Development of severe accident analysis code - Development of a finite element code for lower head failure analysis

    Energy Technology Data Exchange (ETDEWEB)

    Huh, Hoon; Lee, Choong Ho; Choi, Tae Hoon; Kim, Hyun Sup; Kim, Se Ho; Kang, Woo Jong; Seo, Chong Kwan [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

    1995-08-01

    The study concerns the development of analysis models and computer codes for lower head failure analysis when a severe accident occurs in a nuclear reactor system. Although the lower head failure modes consists of several failure modes, the study this year was focused on the global rupture with the collapse pressure and mode by limit analysis and elastic deformation. The behavior of molten core causes elevation of temperature in the reactor vessel wall and deterioration of load-carrying capacity of a reactor vessel. The behavior of molten core and the heat transfer modes were, therefore, postulated in several types and the temperature distributions according to the assumed heat flux modes were calculated. The collapse pressure of a nuclear reactor lower head decreases rapidly with elevation of temperature as time passes. The calculation shows the safety of a nuclear reactor is enhanced with the lager collapse pressure when the hot spot is located far from the pole. 42 refs., 2 tabs., 31 figs. (author)

  17. Use of bayesian operations for diagnosing accidents

    International Nuclear Information System (INIS)

    Kang, K.M.; Jae, M.; Suh, K.Y.

    2005-01-01

    In complex systems, it is necessary to model a logical representation of the overall system interaction with respect to the individual subsystems. Operators are allowed to follow EOPs (Emergency Operating Procedures) when reactor tripped because of accidents. But, it's very difficult to diagnose accidents and find out appropriate procedures to mitigate current accidents in a given short time. Even if they diagnose accidents, it also has possibility to misdiagnose. TMI accident is a good example of operators' errors. Methodology using Influence Diagrams has been developed and applied for representing the dependency behaviors and uncertain behaviors of complex systems. An example to diagnose the accidents such as SLOCA and SGTR with similar symptoms has been introduced. From the constructed model, operators could diagnose accidents at any states of accidents. This model can offer the information about accidents with given symptoms. This model might help operators to diagnose correctly and rapidly. It might be very useful to support operators to reduce human error. Also, from this study, it is applicable to diagnose other accidents with similar symptoms and to analyze causes of reactor trip. (authors)

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

    International Nuclear Information System (INIS)

    Bailly, H.W.

    1988-01-01

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

  19. Modeling to assessment in an environmental consequences of nuclear accidents

    Energy Technology Data Exchange (ETDEWEB)

    Arkhipov, A.N.; Meleshin, A.Yu.; Ponomarev, S.Yu; Ozornov, A.G.; Tepikin, V.E.; Arkhipov, N.P. [Chernobyl Scientific and Technical Centre for International Research, Chernobyl (Ukraine)

    2001-03-01

    To evaluate radiological risk of Chernobyl Exclusion Zone, radionuclide migration mechanisms were investigated including partially vertical migration through soil to plant. After collection and systematization of radiological data on {sup 90}Sr, {sup 137/134}Cs, Ce, and Ru regarding to terrestrial (spatial) and vertical distribution in soil profile in the framework of subproject-2 CheSCIR, the methodology of combining various radio-ecological and geographical information has been developed, which allows to analyze possible schemes of economic use of the contaminated territories. (S. Ohno)

  20. Modeling to assessment in an environmental consequences of nuclear accidents

    International Nuclear Information System (INIS)

    Arkhipov, A.N.; Meleshin, A.Yu.; Ponomarev, S.Yu; Ozornov, A.G.; Tepikin, V.E.; Arkhipov, N.P.

    2001-01-01

    To evaluate radiological risk of Chernobyl Exclusion Zone, radionuclide migration mechanisms were investigated including partially vertical migration through soil to plant. After collection and systematization of radiological data on 90 Sr, 137/134 Cs, Ce, and Ru regarding to terrestrial (spatial) and vertical distribution in soil profile in the framework of subproject-2 CheSCIR, the methodology of combining various radio-ecological and geographical information has been developed, which allows to analyze possible schemes of economic use of the contaminated territories. (S. Ohno)

  1. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Directory of Open Access Journals (Sweden)

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  2. Accident consequence analysis models applied to licensing process of nuclear installations, radioactive and conventional industries

    International Nuclear Information System (INIS)

    Senne Junior, Murillo; Vasconcelos, Vanderley de; Jordao, Elizabete

    2002-01-01

    The industrial accidents happened in the last years, particularly in the eighty's decade, had contributed in a significant way to call the attention to government authorities, industry and society as a whole, demanding mechanisms for preventing episodes that could affect people's safety and environment quality. Techniques and methods already thoroughly used in the nuclear, aeronautic and war industries were then adapted for performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. Some models for analyzing the consequences of accidents involving fire and explosion, used in the licensing processes of nuclear and radioactive facilities, are presented in this paper. These models have also application in the licensing of conventional industrial facilities. (author)

  3. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk

  4. Uncertainty and sensitivity analysis of food pathway results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing-season dose, crop long-term dose, milk growing-season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk

  5. Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

    Energy Technology Data Exchange (ETDEWEB)

    Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

    2016-05-15

    The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

  6. A model of how features of construction projects influence accident occurrence

    OpenAIRE

    Manu, P.

    2017-01-01

    This book chapter in "Valuing People in Construction" (edited by Emuze, F. and Smallwood, J.) presents a study which sought empirical verification of a model of how construction project features (CPFs) influence accident occurrence. A qualitative strategy, in particular phenomenology, involving a range of in-depth interviews with practitioners was used and the findings were subsequently validated using a credibility check involving a survey. Altogether, the findings of the interviews and cred...

  7. Application of multicomponent medium model for numerical simulation of reactor element melting and melt relocation under severe accidents

    International Nuclear Information System (INIS)

    Vladimir Ya Kumaev

    2005-01-01

    Full text of publication follows: Numerical simulation of the melting processes is necessary in substantiating the safety of new generation reactors to determine the quantitative characteristics of the melt formed, destruction of reactor vessel and components, melt interaction processes in the melt localization systems (MLS), formation and transport of hydrogen, radioactive aerosols under severe accidents. The results of computations will be applied in developing the procedures for severe accident management and mitigation of its consequences and designing melt localization systems. The report is devoted to the development and application of the two-dimensional and three-dimensional versions of the DINCOR code intended for numerical simulation of the thermal hydraulic processes in a multicomponent medium with solid-liquid phase changes. The basic set of equations of multicomponent medium is presented. The numerical method to solve the governing equations is discussed. Some examples of two-dimensional code applications are presented. The experience of application of the code has shown that joint calculations of hydrodynamics, heat transfer, stratification and chemical interaction enable the process description accuracy to be significantly increased and the number of initial experimental data to be reduced. The multicomponent medium model can be used as the base for the development of a three-dimensional version of the code. At the same time, it was established that the models being used need be further developed. The most important problems are the following: -development of the local mathematical models of liquefaction and solidification of materials under front melting and melting due to the action of internal sources; -development of the model of incompressible components separation; -development of the models of dissolution and chemical interaction of multicomponent medium components. In conclusion possible verification of the computer code is discussed. (author)

  8. Modelling Reactivity-Initiated-Accident Experiments With Falcon And SCANAIR: A Comparison Exercise

    International Nuclear Information System (INIS)

    Romano, A.; Wallin, H.; Zimmermann, M.A.

    2005-01-01

    A critical assessment is made of the state-of-the-art fuel performance code FALCON in the context of selected Reactivity Initiated Accident (RIA) experiments from the CABRI REP Na series, and contrasts its predictions against those of the extensively benchmarked SCANAIR (Version 3.2) code. The thermal fields in the fuel and cladding, the clad mechanical deformation, and the Fission Gas Release (FGR) are adopted as 'Figures of Merit' by which to judge code performance. Particular attention is paid to the importance of fission-gas-induced clad deformation (which is modelled in SCANAIR, but not in FALCON), relative to that driven by the fuel thermal expansion (which is modelled by both codes). The thermal fields calculated by the codes are in good agreement with each other, especially during the initial stages of the transients --- the adiabatic phase. Larger discrepancies are observed at later times, and are due to the different models applied to calculate the gap conductance. FALCON predicts clad permanent deformations at the end of the transients with a maximum deviation from the experimental measurements of about 20%. Generally, the code always tends to underpredict the measurements. SCANAIR performs similarly, but grossly overpredicts the permanent clad strain for the case involving a very energetic pulse. The fission-gas-driven clad deformation is only relevant for very fast pulse energy injection cases, which are not prototypical of the RIA transients expected in PWRs. The FGR models in FALCON do not capture the mechanism of 'burst-release' in the RIA transients, having been developed for steady-state irradiation conditions. This also explains why they performed poorly when applied to the fast-transient cases analyzed here. In contrast, the FGR results from SCANAIR are in satisfactory agreement with the experimental results. (author)

  9. Modeling of the consequences of an air ingress in the vessel on the source term calculation in a severe accident

    International Nuclear Information System (INIS)

    Beuzet, E.

    2010-01-01

    This work deals with severe accidents in nuclear power plants with an air ingress in the vessel. Air, which is very oxidizing, enhances noticeably ruthenium release, due to a high fuel oxidation. Oxygen affinity is decreasing between zircaloy cladding, fuel and ruthenium which is released from fuel as volatile and radio-toxic oxides. The main objective was to develop models for cladding and fuel oxidation and ruthenium release and implement them in the severe accident code MAAP. For each phenomenon, the literature study contributes to establish numerical models, validated against specific experiments. Cladding oxidation by air leads to a dense zirconia layer and, after cracking, a porous zirconia. Its modelling was validated against QUENCH-10 (KIT) and PARAMETER-SF4 (LUCH) experiments and 'NUREG' oxide mass gain law was selected as the best current compromise. Ruthenium release is modelled as an instantaneous oxide volatilisation at the fuel surface, controlled by fuel oxidation, which is itself based on a surface reaction assumption. These models were validated against a selected set of VERCORS (CEA) experiments. Ruthenium release is well-reproduced, particularly for air and steam atmospheres. To finish, an accidental scenario with air ingress in the vessel was simulated: a first value was given for the time needed to completely release ruthenium still present in core and was compared to the other studies. This simulation underlines the high ruthenium release under air conditions. (author)

  10. Development of intergrated accident management assessment technology; development of interface modules of risk-monitoring system

    Energy Technology Data Exchange (ETDEWEB)

    Kang, S. K.; Park, S. K.; Seok, H.; Kim, D. K.; Han, J. K.; Park, B. R. [KOPEC, Taejeon (Korea)

    2002-03-01

    Based on the development of interface modules with FORTE.- DynaRM can quantify risk model very fast (Very frequent risk model quantification is needed for configuration risk management).- risk monitoring system technology transfer to foreign NPPs. Contribution to component failure and maintenance control automation with the development of Tagging control System. On-Line risk monitoring system development by joint team between Korea Atomic Energy Research Institute and KOPEC is a request by KEPCO. The softwares developed in this study is easily implemented at domestic NPPs without extra study or cost. Economic benefit and Software export to foreign NPPs are expected because of the development of technology related to risk monitoring system and its management. 6 refs., 3 figs., 1 tab. (Author)

  11. Improvement on models associated with LOCA and loss of RHR accidents during shutdown

    International Nuclear Information System (INIS)

    Chang, W. P.; Chung, Y. J.; Kim, W. S.; Kim, K. D.; Lee, S. J.; Jung, J. J.; Ha, G. S.; Son, Y. S.; Chung, B. D.; Han, D. H.; Lee, Y. J.; Hwang, T. S.; Lee, S. Y.; Park, C. Y.; Choi, H. R.; Lee, S. Y.; Choi, J. H.; Ban, C. H.; Bae, G. H.

    1997-07-01

    The characteristics of the best estimate codes available in Korea have been studied through literature surveys for the reliability on LOCA analyses and then, a feasibility study on reduction of capacities of existing safety systems in YGN 3/4 have been carried out using the codes. Since it has been expected to adopt DVI + 4 -Train HPSI in the next generation reactor, the core uncoveries under one DVI line break and 6 cold leg break, which is a requirement for advance d reactor by EPRI, in addition to LBLOCA for reduction effect of SIT capacity, have been analyzed. Finally, an effort on finding the way how the system could be simplified, has been made through the analysis of SIT injection characteristics. On the other hand, the best estimate methodology consisting of uncertainties of the code itself, bias, and application have been developed first and quantification of the uncertainty has been made the case of KORI unit 3 afterward. The prediction capabilities of the best estimate codes and major physical models on the accident under loss of RHR during shutdown have been assessed suing the large scale experimental data delivered from France and then, the assessed codes have been used to provide essential data required for description of operation procedures in YGN 3/4. (author). 64 refs., 45 figs

  12. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale FeCrAl Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Capolungo, L. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wirth, B. D. [Univ. of Tennessee, Knoxville, TN (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced ac- cident tolerance when compared to traditional UO2 fuel zircaloy clad fuel rods. One of the potential replacement claddings are iron-chromium-alunimum (FeCrAl) alloys due to their increased oxidation resistance [1–4] and higher strength [1, 2]. While the oxidation characteristics of FeCrAl are a benefit for accident tolerance, the thermal neu- tron absorption cross section of FeCrAl is about ten times that of Zircaloy. This neutronic penalty necessitates thinner cladding. This allows for slightly larger pellets to give the same cold gap width in the rod. However, the slight increase in pellet diameter is not sufficient to compensate for the neutronic penalty and enriching the fuel beyond the current 5% limit appears to be necessary [5]. Current estimates indicate that this neutronic penalty will impose an increase in fuel cost of 15-35% [1, 2]. In addition to the neutronic disadvantage, it is anticipated that tritium release to the coolant will be larger because the permeability of hydrogen in FeCrAl is about 100 times higher than in Zircaloy [6]. Also, radiation-induced hardening and embrittlement of FeCrAl need to be fully characterized experimentally [7]. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022 [8] multiscale multiphysics modeling approaches have been used to provide insight into these the use of FeCrAl as a cladding material. The purpose of this letter report is to highlight the multiscale modeling effort for iron-chromium-alunimum (FeCrAl) cladding alloys as part of the Nuclear Energy Advanced Modeling and Simulation (NEAMS) program through its Accident Tolerant Fuel (ATF) High Impact Problem (HIP). The approach taken throughout the HIP is to

  13. M5 model tree based predictive modeling of road accidents on non-urban sections of highways in India.

    Science.gov (United States)

    Singh, Gyanendra; Sachdeva, S N; Pal, Mahesh

    2016-11-01

    This work examines the application of M5 model tree and conventionally used fixed/random effect negative binomial (FENB/RENB) regression models for accident prediction on non-urban sections of highway in Haryana (India). Road accident data for a period of 2-6 years on different sections of 8 National and State Highways in Haryana was collected from police records. Data related to road geometry, traffic and road environment related variables was collected through field studies. Total two hundred and twenty two data points were gathered by dividing highways into sections with certain uniform geometric characteristics. For prediction of accident frequencies using fifteen input parameters, two modeling approaches: FENB/RENB regression and M5 model tree were used. Results suggest that both models perform comparably well in terms of correlation coefficient and root mean square error values. M5 model tree provides simple linear equations that are easy to interpret and provide better insight, indicating that this approach can effectively be used as an alternative to RENB approach if the sole purpose is to predict motor vehicle crashes. Sensitivity analysis using M5 model tree also suggests that its results reflect the physical conditions. Both models clearly indicate that to improve safety on Indian highways minor accesses to the highways need to be properly designed and controlled, the service roads to be made functional and dispersion of speeds is to be brought down. Copyright © 2016 Elsevier Ltd. All rights reserved.

  14. Development Status of Accident-tolerant Fuel for Light Water Reactors in Korea

    Directory of Open Access Journals (Sweden)

    Hyun-Gil Kim

    2016-02-01

    Full Text Available For a long time, a top priority in the nuclear industry was the safe, reliable, and economic operation of light water reactors. However, the development of accident-tolerant fuel (ATF became a hot topic in the nuclear research field after the March 2011 events at Fukushima, Japan. In Korea, innovative concepts of ATF have been developing to increase fuel safety and reliability during normal operations, operational transients, and also accident events. The microcell UO2 and high-density composite pellet concepts are being developed as ATF pellets. A microcell UO2 pellet is envisaged to have the enhanced retention capabilities of highly radioactive and corrosive fission products. High-density pellets are expected to be used in combination with the particular ATF cladding concepts. Two concepts—surface-modified Zr-based alloy and SiC composite material—are being developed as ATF cladding, as these innovative concepts can effectively suppress hydrogen explosions and the release of radionuclides into the environment.

  15. Problems of demographic development of the Homel region following the Chernobyl accident

    International Nuclear Information System (INIS)

    Zlotnikov, A.G.

    1993-01-01

    The paper studies the problems of demographic development of the Gomel region linked with the Chernobyl accident and the results of the social polling of the Gomel region population, as well as, of physicians-experts. It is shown that in the Gomel region during the recent years one can observe unprecedented under peace-time conditions reduction of population resulted from the migration of people and from the reduction of the natural increment of population. The results of the opinion polls demonstrated that migration flows of the population are essentially oriented to the out flow. Recently the refugees from the areas of the national conflict come to the Gomel region. 2 tabs

  16. Tools evaluation and development for loss of coolant accidents analysis in research reactors

    International Nuclear Information System (INIS)

    Maprelian, Eduardo; Cabral, Eduardo L.L.; Silva, Antonio T. e

    1999-01-01

    The loss of coolant accidents (LOCA) in pool type research reactors are normally considered as limiting in the licensing process. This paper verifies the viability of the computer code 3D-AIRLOCA to analyze LOCA in a pool type research reactor, and also develops two computer codes LOSS and TEMPLOCA. The computer code LOSS determines the time tom drawn 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. These two coders substitutes the 3D-AIRLOCA in the LOCA analysis for pool type research reactors. (author)

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

    International Nuclear Information System (INIS)

    Jeong, Kwang Il; Lee, Joon Ku

    2016-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2016-05-15

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

  19. Thermal-hydraulic and aerosol containment phenomena modelling in ASTEC severe accident computer code

    International Nuclear Information System (INIS)

    Kljenak, Ivo; Dapper, Maik; Dienstbier, Jiri; Herranz, Luis E.; Koch, Marco K.; Fontanet, Joan

    2010-01-01

    Transients in containment systems of different scales (Phebus.FP containment, KAEVER vessel, Battelle Model Containment, LACE vessel and VVER-1000 nuclear power plant containment) involving thermal-hydraulic phenomena and aerosol behaviour, were simulated with the computer integral code ASTEC. The results of the simulations in the first four facilities were compared with experimental results, whereas the results of the simulated accident in the VVER-1000 containment were compared to results obtained with the MELCOR code. The main purpose of the simulations was the validation of the CPA module of the ASTEC code. The calculated results support the applicability of the code for predicting in-containment thermal-hydraulic and aerosol phenomena during a severe accident in a nuclear power plant.

  20. A dynamic model for the study of evacuation under accident conditions

    International Nuclear Information System (INIS)

    Boeri, G.C.; Caracciolo, R.; Sepede, M.; Casiroli, F.; Rodriguez, R.

    1987-01-01

    Information techniques and models are being used to simulate the evacuation of people living around a nuclear power plant and these methods are being increasingly used for planning purposes. In this study vehicular mobility on a complex road network following an accident has been considered and applied to the specifications of the emergency plans. Simulation tests of the mobility relevant to different combinations of time and meteorological conditions have been undertaken through use of the computer code TRIPS in order to assess the impact on the road network of an accident situation and the preparedness measures. The study has led to a description of the accessibility time curves related to the population gathering centres and has enabled identification of the best routes in order to achieve the minimum travel time. (author)

  1. Modelling road accident blackspots data with the discrete generalized Pareto distribution.

    Science.gov (United States)

    Prieto, Faustino; Gómez-Déniz, Emilio; Sarabia, José María

    2014-10-01

    This study shows how road traffic networks events, in particular road accidents on blackspots, can be modelled with simple probabilistic distributions. We considered the number of crashes and the number of fatalities on Spanish blackspots in the period 2003-2007, from Spanish General Directorate of Traffic (DGT). We modelled those datasets, respectively, with the discrete generalized Pareto distribution (a discrete parametric model with three parameters) and with the discrete Lomax distribution (a discrete parametric model with two parameters, and particular case of the previous model). For that, we analyzed the basic properties of both parametric models: cumulative distribution, survival, probability mass, quantile and hazard functions, genesis and rth-order moments; applied two estimation methods of their parameters: the μ and (μ+1) frequency method and the maximum likelihood method; used two goodness-of-fit tests: Chi-square test and discrete Kolmogorov-Smirnov test based on bootstrap resampling; and compared them with the classical negative binomial distribution in terms of absolute probabilities and in models including covariates. We found that those probabilistic models can be useful to describe the road accident blackspots datasets analyzed. Copyright © 2014 Elsevier Ltd. All rights reserved.

  2. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    Energy Technology Data Exchange (ETDEWEB)

    Kao, S.P.; Chang, S.K.; Huang, H.C. [Nuclear Training Branch, Northeast Utilities, Waterford, CT (United States)

    1995-09-01

    The replacement of the NSSS (Nuclear Steam Supply System) model on the Millstone 2 full-scope simulator has significantly increased its fidelity to simulate adverse conditions in the RCS. The new simulator NSSS model is a real-time derivative of the Nuclear Plant Analyzer by ABB. The thermal-hydraulic model is a five-equation, non-homogeneous model for water, steam, and non-condensible gases. The neutronic model is a three-dimensional nodal diffusion model. In order to certify the new NSSS model for operator training, an extensive validation effort has been performed by benchmarking the model performance against RELAP5/MOD2. This paper presents the validation results for the cases of small-and large-break loss-of-coolant accidents (LOCA). Detailed comparisons in the phenomena of reflux-condensation, phase separation, and two-phase natural circulation are discussed.

  3. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    International Nuclear Information System (INIS)

    Robbe, M.F.

    2003-12-01

    Two kinds of accidents, liable to occur in the primary circuit of a Pressurized Water Reactor and involving fast dynamic phenomena, are analyzed. The Loss Of Coolant Accident (LOCA) is the accident used to define the current PWR. It consists in a large-size break located in a pipe of the primary circuit. A blowdown wave propagates through the circuit. The pressure differences between the different zones of the reactor induce high stresses in the structures of the lower head and may degrade the reactor core. The primary circuit starts emptying from the break opening. Pressure decreases very quickly, involving a large steaming. Two thermal-hydraulic simulations of the blowdown phase of a LOCA are computed with the Europlexus code. The primary circuit is represented by a pipe-model including the hydraulic peculiarities of the circuit. The main differences between both computations concern the kind of reactor, the break location and model, and the initialization of the accidental operation. Steam explosion is a hypothetical severe accident liable to happen after a core melting. The molten part of the core (called corium) falls in the lower part of the reactor. The interaction between the hot corium and the cold water remaining at the bottom of the vessel induces a massive and violent vaporization of water, similar to an explosive phenomenon. A shock wave propagates in the vessel. what can damage seriously the neighbouring structures or drill the vessel. This work presents a synthesis of in-vessel parametrical studies carried out with the Europlexus code, the linkage of the thermal-hydraulic code Mc3d dedicated to the pre-mixing phase with the Europlexus code dealing with the explosion, and finally a benchmark between the Cigalon and Europlexus codes relative to the Vulcano mock-up. (author)

  4. Estimation of expected number of accidents and workforce unavailability through Bayesian population variability analysis and Markov-based model

    International Nuclear Information System (INIS)

    Chagas Moura, Márcio das; Azevedo, Rafael Valença; Droguett, Enrique López; Chaves, Leandro Rego; Lins, Isis Didier

    2016-01-01

    Occupational accidents pose several negative consequences to employees, employers, environment and people surrounding the locale where the accident takes place. Some types of accidents correspond to low frequency-high consequence (long sick leaves) events, and then classical statistical approaches are ineffective in these cases because the available dataset is generally sparse and contain censored recordings. In this context, we propose a Bayesian population variability method for the estimation of the distributions of the rates of accident and recovery. Given these distributions, a Markov-based model will be used to estimate the uncertainty over the expected number of accidents and the work time loss. Thus, the use of Bayesian analysis along with the Markov approach aims at investigating future trends regarding occupational accidents in a workplace as well as enabling a better management of the labor force and prevention efforts. One application example is presented in order to validate the proposed approach; this case uses available data gathered from a hydropower company in Brazil. - Highlights: • This paper proposes a Bayesian method to estimate rates of accident and recovery. • The model requires simple data likely to be available in the company database. • These results show the proposed model is not too sensitive to the prior estimates.

  5. Nuclear Energy Advanced Modeling and Simulation (NEAMS) Accident Tolerant Fuels High Impact Problem: Coordinate Multiscale U3Si2 Modeling

    Energy Technology Data Exchange (ETDEWEB)

    Gamble, K. A. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hales, J. D. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Miao, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2017-07-26

    Since the events at the Fukushima-Daiichi nuclear power plant in March 2011 significant research has unfolded at national laboratories, universities and other institutions into alternative materials that have potential enhanced accident tolerance when compared to traditional \\uo~fuel zircaloy clad fuel rods. One of the potential replacement fuels is uranium silicide (\\usi) for its higher thermal conductivity and uranium density. The lower melting temperature is of potential concern during postulated accident conditions. Another disadvantage for \\usi~ is the lack of experimental data under power reactor conditions. Due to the aggressive development schedule for inserting some of the potential materials into lead test assemblies or rods by 2022~\\cite{bragg-sitton_2014} multiscale multiphysics modeling approaches have been used to provide insight into these materials. \\\\ \

  6. Model review and evaluation for application in DOE safety basis documentation of chemical accidents - modeling guidance for atmospheric dispersion and consequence assessment

    Energy Technology Data Exchange (ETDEWEB)

    Lazaro, M. A. [Argonne National Lab. (ANL), Argonne, IL (United States); Woodarad, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Hanna, S. R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hesse, D. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Huang, J. -C. [Argonne National Lab. (ANL), Argonne, IL (United States); Lewis, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Mazzola, C. A. [Argonne National Lab. (ANL), Argonne, IL (United States)

    1997-09-01

    The U.S. Department of Energy (DOE), through its Defense Programs (DP), Office of Engineering and Operations Suppon, established the Accident Phenomenology and Consequence (AP AC) Methodology Evaluation Program to identify and evaluate methodologies and computer codes to support accident phenomenological and consequence calculations for both radiological and nonradiological materials at DOE facilities and to identify development needs. The program is also intended to define and recommend "best or good engineering/safety analysis practices" to be followed in preparing ''design or beyond design basis" assessments to be included in DOE nuclear and nonnuclear facility safety documents. The AP AC effort is intended to provide scientifically sound and more consistent analytical approaches, by identifying model selection procedures and application methodologies, in order to enhance safety analysis activities throughout the DOE complex.

  7. A strategy to the development of a human error analysis method for accident management in nuclear power plants using industrial accident dynamics

    International Nuclear Information System (INIS)

    Lee, Yong Hee; Kim, Jae Whan; Jung, Won Dae; Ha, Jae Ju

    1998-06-01

    This technical report describes the early progress of he establishment of a human error analysis method as a part of a human reliability analysis(HRA) method for the assessment of the human error potential in a given accident management strategy. At first, we review the shortages and limitations of the existing HRA methods through an example application. In order to enhance the bias to the quantitative aspect of the HRA method, we focused to the qualitative aspect, i.e., human error analysis(HEA), during the proposition of a strategy to the new method. For the establishment of a new HEA method, we discuss the basic theories and approaches to the human error in industry, and propose three basic requirements that should be maintained as pre-requisites for HEA method in practice. Finally, we test IAD(Industrial Accident Dynamics) which has been widely utilized in industrial fields, in order to know whether IAD can be so easily modified and extended to the nuclear power plant applications. We try to apply IAD to the same example case and develop new taxonomy of the performance shaping factors in accident management and their influence matrix, which could enhance the IAD method as an HEA method. (author). 33 refs., 17 tabs., 20 figs

  8. Modelling and assessment of accident consequences: Development of a computer-assisted decision-support system RODOS/RESY for nuclear emergencies; Modellierung und Abschaetzung von Unfallfolgen: Entwicklung des rechnergestuetzen Entscheidungshilfesystems RODOS/RESY fuer kerntechnische Notfaelle

    Energy Technology Data Exchange (ETDEWEB)

    Benz, G. [Dr. Trippe Ingenieurgesellschaft mbH, Karlsruhe (Germany); Ehrhardt, J. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Faude, D. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Fischer, F. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Hasemann, I. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Lorenz, A. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Meyer, D. [Dr. Trippe Ingenieurgesellschaft mbH, Karlsruhe (Germany); Paesler-Sauer, J. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Rafat, M. [Dr. Trippe Ingenieurgesellschaft mbH, Karlsruhe (Germany); Schichtel, T. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Schuele, O. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik; Steinhauer, C. [Forschungszentrum Karlsruhe GmbH Technik und Umwelt (Germany). Inst. fuer Neutronenphysik und Reaktortechnik

    1995-08-01

    In cooperation with NRPB, the specifications of the mainframe COSYMA version 95/1 and the PC COSYMA version 2.0 were prepared and the corresponding modifications implemented. Important improvements are dose-rate dependent models for deterministic health effects, the time dependent efficiency of stable iodine tablets, the extension of data bases for the inclusion of activation products, and supplementary evaluation programs. PC COSYMA has been completed by an economics module, further options in the ingestion pathways, and a graphics package for presenting assessment results. COSYMA has been applied for probabilistic dose assessments within paramter studies and special investigations of EPR concepts. RODOS, the real-time on-line decision support system for nuclear emergency management, has been further developed with the aim of the first pilot version 2.0 for pre-operational application in the second half of 1995. At present, some 20 institutes in the EU, 8 institutes in Russia, Belarus and Ukraine, and 5 institutes in central and eastern European countries cooperate with FZKA/INR as main coordinator. The working meeting of all contractors in Interlaken (CH), 20-24 June 1994, lead to the consolidation of the international cooperation and a broad consensus about the future role and use of the RODOS-system, in particular in the Eastern countries. At the national level, the approval of a strategy and position paper on RODOS/RESY (the near range and early phase subsystem) by subcommittees of the German Radiation Protection Commission can be considered as a step further to the future use of the system in emergency centres of the Federal States. (orig./HP)

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

    International Nuclear Information System (INIS)

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

    2016-09-01

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

  10. Effects of neutron streaming and geometric models on molten fuel recriticality accidents

    International Nuclear Information System (INIS)

    McLaughlin, T.P.

    1975-10-01

    A postulated fast reactor accident which has been extant for many years is a recriticality following partial or complete core melting. Independently of the cause or probability of such a situation, certain cases can be defined and some facets of the dynamic history of these cases can be described with more than enough accuracy for safety considerations. Calculations were made with the PAD code for systems with 10 vol percent voids and varying reactivity insertion rates. Additionally, two distinct geometric and equation of state models were investigated in conjunction with a model which accounted for possible neutron streaming reactivity effects. Significant results include fission and kinetic energy, temperatures and pressures

  11. Inclusion of models to describe severe accident conditions in the fuel simulation code DIONISIO

    Energy Technology Data Exchange (ETDEWEB)

    Lemes, Martín; Soba, Alejandro [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Daverio, Hernando [Gerencia Reactores y Centrales Nucleares, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina); Denis, Alicia [Sección Códigos y Modelos, Gerencia Ciclo del Combustible Nuclear, Comisión Nacional de Energía Atómica, Avenida General Paz 1499, 1650 San Martín, Provincia de Buenos Aires (Argentina)

    2017-04-15

    The simulation of fuel rod behavior is a complex task that demands not only accurate models to describe the numerous phenomena occurring in the pellet, cladding and internal rod atmosphere but also an adequate interconnection between them. In the last years several models have been incorporated to the DIONISIO code with the purpose of increasing its precision and reliability. After the regrettable events at Fukushima, the need for codes capable of simulating nuclear fuels under accident conditions has come forth. Heat removal occurs in a quite different way than during normal operation and this fact determines a completely new set of conditions for the fuel materials. A detailed description of the different regimes the coolant may exhibit in such a wide variety of scenarios requires a thermal-hydraulic formulation not suitable to be included in a fuel performance code. Moreover, there exist a number of reliable and famous codes that perform this task. Nevertheless, and keeping in mind the purpose of building a code focused on the fuel behavior, a subroutine was developed for the DIONISIO code that performs a simplified analysis of the coolant in a PWR, restricted to the more representative situations and provides to the fuel simulation the boundary conditions necessary to reproduce accidental situations. In the present work this subroutine is described and the results of different comparisons with experimental data and with thermal-hydraulic codes are offered. It is verified that, in spite of its comparative simplicity, the predictions of this module of DIONISIO do not differ significantly from those of the specific, complex codes.

  12. Comparison of the foodchain transport models of WASH-1400 and MARC using the accident consequence model UFOMOD

    International Nuclear Information System (INIS)

    Steinhauer, C.

    1985-04-01

    Within the frame of the contract with the European Community 'Methods for Assessing the Radiological Impact of Accidents' (CEC-MARIA) comparative accident consequence assessments were performed with the computer code UFOMOD, replacing the currently implemented foodchain transport model of the WASH-1400 study by the dynamic transport model of the MARC methodology. The calculations were based on the release category FK2 of the German Risk Study with meteorological data representing four different regions of the Federal Republic of Germany. The study of seasonal variations was carried out with the MARC data for four representative times of deposition with an agricultural practice adopted in the UK. In this report the differences are presented which are observed in the potential doses due to ingestion, the areas affected by food-bans and the late health effects when using both models and taking the influence of seasonal effects into account. (orig.) [de

  13. Development of high-performance monitoring system under severe accident condition

    International Nuclear Information System (INIS)

    Takeuchi, Tomoaki; Tsuchiya, Kunihiro; Ishihara, Masahiro; Komanome, H.; Miura, K.

    2017-01-01

    A research and development of a monitoring system for NPPs situations even during severe accidents have been performed. The R and D consists of the three objectives. The major findings are briefly summarized in the followings: 1) Radiation-resistant monitoring camera. The image sensor with the photogate and three transistors was found to be advantageous in terms of dark current and sensitivity. In addition, radiation-resistant optical parts and signal circuits were successfully fabricated. The results suggested that the monitoring camera system with 10 6 Gy in radiation resistance was possible. 2) Radiation-resistant in-water wireless transmission system. A two-dimensional LED matrix with 10 6 Gy in radiation resistance and a camera were used as the transmission devices. The results of the in-water transmission tests suggested that stable wireless transmission between 5 m distance was possible even with bubble, turbidity, or obstacles. 3) Heat-resistant signal cable. In order to develop a cable that can transmit the data inside reactor pressure vessels, heat-proof tests were performed for candidate metallic sheath materials of mineral insulation (MI) cables. The results indicated MI cables which can be used at 1000degC in air were possible. These results indicate the feasibility of the monitoring system even during severe accidents. (author)

  14. Development of training system to prevent accidents during decommissioning of nuclear facilities

    International Nuclear Information System (INIS)

    Jeong, Kwanseong; Moon, Jeikwon; Choi, Byungseon; Hyun, Dongjun; Lee, Jonghwan; Kim, Ikjune; Kim, Geunho; Seo, Jaeseok

    2014-01-01

    Decommissioning workers need familiarization with working environments because working environment is under high radioactivity and work difficulty during decommissioning of nuclear facilities. On-the-job training of decommissioning works could effectively train decommissioning workers but this training approach could consume much costs and poor modifications of scenarios. The efficiency of virtual training system could be much better than that of physical training system. This paper was intended to develop the training system to prevent accidents for decommissioning of nuclear facilities. The requirements for the training system were drawn. The data management modules for the training system were designed. The training system of decommissioning workers was developed on the basis of virtual reality which is flexibly modified. The visualization and measurement in the training system were real-time done according as changes of the decommissioning scenario. It can be concluded that this training system enables the subject to improve his familiarization about working environments and to prevent accidents during decommissioning of nuclear facilities

  15. Development of Cold Spray Coatings for Accident-Tolerant Fuel Cladding in Light Water Reactors

    Science.gov (United States)

    Maier, Benjamin; Yeom, Hwasung; Johnson, Greg; Dabney, Tyler; Walters, Jorie; Romero, Javier; Shah, Hemant; Xu, Peng; Sridharan, Kumar

    2018-02-01

    The cold spray coating process has been developed at the University of Wisconsin-Madison for the deposition of oxidation-resistant coatings on zirconium alloy light water reactor fuel cladding with the goal of improving accident tolerance during loss of coolant scenarios. Coatings of metallic (Cr), alloy (FeCrAl), and ceramic (Ti2AlC) materials were successfully deposited on zirconium alloy flats and cladding tube sections by optimizing the powder size, gas preheat temperature, pressure and composition, and other process parameters. The coatings were dense and exhibited excellent adhesion to the substrate. Evaluation of the samples after high-temperature oxidation tests at temperatures up to 1300°C showed that the cold spray coatings significantly mitigate oxidation kinetics because of the formation of thin passive oxide layers on the surface. The results of the study indicate that the cold spray coating process is a viable near-term option for developing accident-tolerant zirconium alloy fuel cladding.

  16. Development of training system to prevent accidents during decommissioning of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Kwanseong; Moon, Jeikwon; Choi, Byungseon; Hyun, Dongjun; Lee, Jonghwan; Kim, Ikjune; Kim, Geunho; Seo, Jaeseok [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-05-15

    Decommissioning workers need familiarization with working environments because working environment is under high radioactivity and work difficulty during decommissioning of nuclear facilities. On-the-job training of decommissioning works could effectively train decommissioning workers but this training approach could consume much costs and poor modifications of scenarios. The efficiency of virtual training system could be much better than that of physical training system. This paper was intended to develop the training system to prevent accidents for decommissioning of nuclear facilities. The requirements for the training system were drawn. The data management modules for the training system were designed. The training system of decommissioning workers was developed on the basis of virtual reality which is flexibly modified. The visualization and measurement in the training system were real-time done according as changes of the decommissioning scenario. It can be concluded that this training system enables the subject to improve his familiarization about working environments and to prevent accidents during decommissioning of nuclear facilities.

  17. AN ANALYSIS OF ACCIDENT TRENDS AND MODELING OF SAFETY INDICES IN AN INDIAN CONSTRUCTION ORGANIZATION

    Directory of Open Access Journals (Sweden)

    Sunku Venkata Siva Rajaprasad

    2016-09-01

    Full Text Available Construction industry has been recognized as a hazardous industry in many countries due to distinct nature of execution of works.The accident rate in construction sector is high all over the world due to dynamic nature of work activities. Occurrence of accidents and its severity in construction industry is several times higher than the manufacturing industries. The study was limited to a major construction organization in India to examine the trends in construction accidents for the period 2008-2014. In India, safety performance is gauged basing on safety indices; frequency, severity and incidence rates. It is not practicable to take decisions or to implement safety strategies on the basis of indices. The data used for this study was collected from a leading construction organization involved in execution of major construction activities all over India and abroad. The multiple regression method was adopted to model the pattern of safety indices wise .The pattern showed that significant relationships exist between the three safety indices and the related independent variables.

  18. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion

  19. The Application of Data Mining Technology to Build a Forecasting Model for Classification of Road Traffic Accidents

    Directory of Open Access Journals (Sweden)

    Yau-Ren Shiau

    2015-01-01

    Full Text Available With the ever-increasing number of vehicles on the road, traffic accidents have also increased, resulting in the loss of lives and properties, as well as immeasurable social costs. The environment, time, and region influence the occurrence of traffic accidents. The life and property loss is expected to be reduced by improving traffic engineering, education, and administration of law and advocacy. This study observed 2,471 traffic accidents which occurred in central Taiwan from January to December 2011 and used the Recursive Feature Elimination (RFE of Feature Selection to screen the important factors affecting traffic accidents. It then established models to analyze traffic accidents with various methods, such as Fuzzy Robust Principal Component Analysis (FRPCA, Backpropagation Neural Network (BPNN, and Logistic Regression (LR. The proposed model aims to probe into the environments of traffic accidents, as well as the relationships between the variables of road designs, rule-violation items, and accident types. The results showed that the accuracy rate of classifiers FRPCA-BPNN (85.89% and FRPCA-LR (85.14% combined with FRPCA is higher than that of BPNN (84.37% and LR (85.06% by 1.52% and 0.08%, respectively. Moreover, the performance of FRPCA-BPNN and FRPCA-LR combined with FRPCA in classification prediction is better than that of BPNN and LR.

  20. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, water ingestion dose, milk growing season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meat, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of 1-131 at which disposal of crops will be initiated due to accidents that occur during the growing season

  1. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    Helton, J.C; Johnson, J.D; Rollstin, J.A; Shiver, A.W; Sprung, J.L

    1995-01-01

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing-season dose, crop long-term dose, water ingestion dose, milk growing-season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meet, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of I-131 at which disposal of crops will be initiated due to accidents that occur during the growing season. Reducing the uncertainty in the preceding variables was found to substantially reduce the uncertainty in the

  2. Radionuclides from the Fukushima accident in the air over Lithuania: measurement and modelling approaches.

    Science.gov (United States)

    Lujanienė, G; Byčenkienė, S; Povinec, P P; Gera, M

    2012-12-01

    Analyses of (131)I, (137)Cs and (134)Cs in airborne aerosols were carried out in daily samples in Vilnius, Lithuania after the Fukushima accident during the period of March-April, 2011. The activity concentrations of (131)I and (137)Cs ranged from 12 μBq/m(3) and 1.4 μBq/m(3) to 3700 μBq/m(3) and 1040 μBq/m(3), respectively. The activity concentration of (239,240)Pu in one aerosol sample collected from 23 March to 15 April, 2011 was found to be 44.5 nBq/m(3). The two maxima found in radionuclide concentrations were related to complicated long-range air mass transport from Japan across the Pacific, the North America and the Atlantic Ocean to Central Europe as indicated by modelling. HYSPLIT backward trajectories and meteorological data were applied for interpretation of activity variations of measured radionuclides observed at the site of investigation. (7)Be and (212)Pb activity concentrations and their ratios were used as tracers of vertical transport of air masses. Fukushima data were compared with the data obtained during the Chernobyl accident and in the post Chernobyl period. The activity concentrations of (131)I and (137)Cs were found to be by 4 orders of magnitude lower as compared to the Chernobyl accident. The activity ratio of (134)Cs/(137)Cs was around 1 with small variations only. The activity ratio of (238)Pu/(239,240)Pu in the aerosol sample was 1.2, indicating a presence of the spent fuel of different origin than that of the Chernobyl accident. Copyright © 2011 Elsevier Ltd. All rights reserved.

  3. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    2004-01-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

  4. A microcomputer-based model for identifying urban and suburban roadways with critical large truck accident rates

    International Nuclear Information System (INIS)

    Brogan, J.D.; Cashwell, J.W.

    1992-01-01

    This paper presents an overview of techniques for merging highway accident record and roadway inventory files and employing the combined data set to identify spots or sections on highway facilities in urban and suburban areas with unusually high large truck accident rates. A statistical technique, the rate/quality control method, is used to calculate a critical rate for each location of interest. This critical rate may then be compared to the location's actual accident rate to identify locations for further study. Model enhancements and modifications are described to enable the technique to be employed in the evaluation of routing alternatives for the transport of radioactive material

  5. Modelling of HTR Confinement Behaviour during Accidents Involving Breach of the Helium Pressure Boundary

    Directory of Open Access Journals (Sweden)

    Joan Fontanet

    2009-01-01

    Full Text Available Development of HTRs requires the performance of a thorough safety study, which includes accident analyses. Confinement building performance is a key element of the system since the behaviour of aerosol and attached fission products within the building is of an utmost relevance in terms of the potential source term to the environment. This paper explores the available simulation capabilities (ASTEC and CONTAIN codes and illustrates the performance of a postulated HTR vented confinement under prototypical accident conditions by a scoping study based on two accident sequences characterized by Helium Pressure Boundary breaches, a small and a large break. The results obtained indicate that both codes predict very similar thermal-hydraulic responses of the confinement both in magnitude and timing. As for the aerosol behaviour, both codes predict that most of the inventory coming into the confinement is eventually depleted on the walls and only about 1% of the aerosol dust is released to the environment. The crosscomparison of codes states that largest differences are in the intercompartmental flows and the in-compartment gas composition.

  6. Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment computer code under high burnup condition

    International Nuclear Information System (INIS)

    Kim, H. D.; Kim, D. H.; Park, S. Y.; Park, J. H.

    2005-10-01

    This project was performed by KAERI in the frame of construction of the international cooperative basis on the nuclear energy. This was supported from MOST under the title of 'Establishment of Technical Collaboration basis between Korea and France for the development of severe accident assessment