WorldWideScience

Sample records for accident development modeling

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

    Energy Technology Data Exchange (ETDEWEB)

    Lui, C.H. [Nuclear Regulatory Commission, Washington, DC (United States); Galyean, W.J.; Brownson, D.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

    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. 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.).

  3. Development and application of traffic accident density estimation models using kernel density estimation

    OpenAIRE

    Seiji Hashimoto; Syuji Yoshiki; Ryoko Saeki; Yasuhiro Mimura; Ryosuke Ando; Shutaro Nanba

    2016-01-01

    Traffic accident frequency has been decreasing in Japan in recent years. Nevertheless, many accidents still occur on residential roads. Area-wide traffic calming measures including Zone 30, which discourages traffic by setting a speed limit of 30 km/h in residential areas, have been implemented. However, no objective implementation method has been established. Development of a model for traffic accident density estimation explained by GIS data can enable the determination of dangerous areas o...

  4. Development of MAAP5.0.3 Spent Fuel Pool Model for Severe Accident Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    SFP model for OPR100 type NPP developed using MAAP 5.0.3. As expected, the accident progression in the SFP is very slow if the SFP integrity in maintained.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1996-06-01

    The study supports the development project of STUK on `Living` PSA Level 2. The main work objective is to develop review tools for the Level 2 PSA studies underway at the utilities. The SPSA (STUK PSA) code is specifically designed for the purpose. In this work, SPSA is utilized as the Level 2 programming and calculation tool. A containment event tree (CET) model is built for analysis of severe accidents at the Loviisa pressurized water reactor (PWR) 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 include new research results, and so it facilitates the Living PSA concept on Level 2 as well. The analyses of the study are limited to severe accidents starting from full-power operation and leading to core melting at a low primary system pressure. Severe accident progression from five plant damage states (PDSs) is examined, however the integration with Level 1 is deferred to more definitive, integrated, safety assessments. (34 refs., 5 figs., 9 tabs.).

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

    International Nuclear Information System (INIS)

    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

  7. Accident consequence assessment code development

    International Nuclear Information System (INIS)

    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)

  8. Development of an Ontology to Assist the Modeling of Accident Scenarii "Application on Railroad Transport "

    CERN Document Server

    Maalel, Ahmed; Mejri, Lassad; Ghezela, Henda Hajjami Ben

    2012-01-01

    In a world where communication and information sharing are at the heart of our business, the terminology needs are most pressing. It has become imperative to identify the terms used and defined in a consensual and coherent way while preserving linguistic diversity. To streamline and strengthen the process of acquisition, representation and exploitation of scenarii of train accidents, it is necessary to harmonize and standardize the terminology used by players in the security field. The research aims to significantly improve analytical activities and operations of the various safety studies, by tracking the error in system, hardware, software and human. This paper presents the contribution of ontology to modeling scenarii for rail accidents through a knowledge model based on a generic ontology and domain ontology. After a detailed presentation of the state of the art material, this article presents the first results of the developed model.

  9. Development of a Gravid Uterus Model for the Study of Road Accidents Involving Pregnant Women.

    Science.gov (United States)

    Auriault, F; Thollon, L; Behr, M

    2016-01-01

    Car accident simulations involving pregnant women are well documented in the literature and suggest that intra-uterine pressure could be responsible for the phenomenon of placental abruption, underlining the need for a realistic amniotic fluid model, including fluid-structure interactions (FSI). This study reports the development and validation of an amniotic fluid model using an Arbitrary Lagrangian Eulerian formulation in the LS-DYNA environment. Dedicated to the study of the mechanisms responsible for fetal injuries resulting from road accidents, the fluid model was validated using dynamic loading tests. Drop tests were performed on a deformable water-filled container at acceleration levels that would be experienced in a gravid uterus during a frontal car collision at 25 kph. During the test device braking phase, container deformation induced by inertial effects and FSI was recorded by kinematic analysis. These tests were then simulated in the LS-DYNA environment to validate a fluid model under dynamic loading, based on the container deformations. Finally, the coupling between the amniotic fluid model and an existing finite-element full-body pregnant woman model was validated in terms of pressure. To do so, experimental test results performed on four postmortem human surrogates (PMHS) (in which a physical gravid uterus model was inserted) were used. The experimental intra-uterine pressure from these tests was compared to intra uterine pressure from a numerical simulation performed under the same loading conditions. Both free fall numerical and experimental responses appear strongly correlated. The relationship between the amniotic fluid model and pregnant woman model provide intra-uterine pressure values correlated with the experimental test responses. The use of an Arbitrary Lagrangian Eulerian formulation allows the analysis of FSI between the amniotic fluid and the gravid uterus during a road accident involving pregnant women. PMID:26592419

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

    International Nuclear Information System (INIS)

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

  11. Developments in Reactor and Economic Modelling Considering the Performance of Accident Tolerant Fuels

    International Nuclear Information System (INIS)

    Accident tolerant fuel (ATF) technology is being developed to enhance the safety performance of nuclear fuels and cladding. The development and testing of ATF materials by NNL through its Nuclear Fuel Centre of Excellence is being complemented by parallel developments in fuel performance modelling, in addition to reactor physics and economic calculations to optimise ATF fuel. An approach for preliminary optimisation of ATF fuel pin and cladding parameters, in typical commercial PWRs is described, including an initial optimisation of uranium nitride (UN) fuel pellet dimensions and enrichment (combined with zirconium cladding) and for silicon carbide composite (SiC) clad fuel (combined with uranium oxide fuel (UO2)). In order to optimise pin reactivity, pellet diameter is less for UN compared with UO2. A lower feed enrichment was required to give an equilibrium energy output close to the equivalent UO2 fuel. Modelling this design indicates that there is a potential economic benefit, through lower fuel assembly costs, when using optimised UN fuel compared with standard UO2 PWR fuel. For standard UO2 fuel, full core calculations have examined the reactivity benefit when replacing zircaloy clad for SiC. Calculations assume idealised SiC clad thicknesses similar to those used with current zircaloy clads. An economic analysis, considering current cost estimates of SiC clad manufacture, indicates SiC clad fuel assembly costs are significantly increased. However, there remains scope for offsetting these increased fuel costs through optimised reactor operation by taking advantage of the reduced parasitic neutron absorption or higher temperature tolerance of SiC clad. An initial assessment is also undertaken of how the performance of the higher density uranium nitride fuel compares against key PWR safety measures: considering pin power peaking, shutdown margin, moderator temperature coefficients, boron reactivity worth, delayed neutron fractions and boration limits. All

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

    International Nuclear Information System (INIS)

    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

  13. Development and application of a random walk model of atmospheric diffusion in the emergency response of nuclear accidents

    Institute of Scientific and Technical Information of China (English)

    CHI Bing; LI Hong; FANG Dong

    2007-01-01

    Plume concentration prediction is one of the main contents of radioactive consequence assessment for early emergency response to nuclear accidents. Random characteristics of atmospheric diffusion itself was described, a random walk model of atmospheric diffusion (Random Walk) was introduced and compared with the Lagrangian puff model (RIMPUFF) in the nuclear emergency decision support system (RODOS) developed by the European Community for verification. The results show the concentrations calculated by the two models are quite close except that the plume area calculated by Random Walk is a little smaller than that by RIMPUFF. The random walk model for atmospheric diffusion can simulate the atmospheric diffusion in case of nuclear accidents, and provide more actual information for early emergency and consequence assessment as one of the atmospheric diffusion module of the nuclear emergency decision support system.

  14. Review of models applicable to accident aerosols

    International Nuclear Information System (INIS)

    Estimations of potential airborne-particle releases are essential in safety assessments of nuclear-fuel facilities. This report is a review of aerosol behavior models that have potential applications for predicting aerosol characteristics in compartments containing accident-generated aerosol sources. Such characterization of the accident-generated aerosols is a necessary step toward estimating their eventual release in any accident scenario. Existing aerosol models can predict the size distribution, concentration, and composition of aerosols as they are acted on by ventilation, diffusion, gravity, coagulation, and other phenomena. Models developed in the fields of fluid mechanics, indoor air pollution, and nuclear-reactor accidents are reviewed with this nuclear fuel facility application in mind. The various capabilities of modeling aerosol behavior are tabulated and discussed, and recommendations are made for applying the models to problems of differing complexity

  15. Development of comprehensive accident models for two-lane rural highways using exposure, geometry, consistency and context variables.

    Science.gov (United States)

    Cafiso, Salvatore; Di Graziano, Alessandro; Di Silvestro, Giacomo; La Cava, Grazia; Persaud, Bhagwant

    2010-07-01

    In Europe, approximately 60% of road accident fatalities occur on two-lane rural roads. Thus, research to develop and enhance explanatory and predictive models for this road type continues to be of interest in mitigating these accidents. To this end, this paper describes a novel and extensive data collection and modeling effort to define accident models for two-lane road sections based on a unique combination of exposure, geometry, consistency and context variables directly related to the safety performance. The first part of the paper documents how these were identified for the segmentation of highways into homogeneous sections. Next, is a description of the extensive data collection effort that utilized differential cinematic GPS surveys to define the horizontal alignment variables, and road safety inspections (RSIs) to quantify the other road characteristics related to safety. The final part of the paper focuses on the calibration of models for estimating the expected number of accidents on homogeneous sections that can be characterized by constant values of the explanatory variables. Several candidate models were considered for calibration using the Generalized Linear Modeling (GLM) approach. After considering the statistical significance of the parameters related to exposure, geometry, consistency and context factors, and goodness of fit statistics, 19 models were ranked and three were selected as the recommended models. The first of the three is a base model, with length and traffic as the only predictor variables; since these variables are the only ones likely to be available network-wide, this base model can be used in an empirical Bayesian calculation to conduct network screening for ranking "sites with promise" of safety improvement. The other two models represent the best statistical fits with different combinations of significant variables related to exposure, geometry, consistency and context factors. These multiple variable models can be used, with

  16. Modeling accidents for prioritizing prevention

    International Nuclear Information System (INIS)

    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

  17. 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. PMID:26687540

  18. 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.

  19. Development of TRAIN for accident management

    International Nuclear Information System (INIS)

    Severe accident management can be defined as the use of existing and alternative resources, systems, and actions to prevent or mitigate a core-melt accident in nuclear power plants. TRAIN (Training pRogram for AMP In NPP), developed for training control room staff and the technical group, is introduced in this paper. The TRAIN composes of phenomenological knowledge base (KB), accident sequence KB and accident management procedures with AM strategy control diagrams and information needs. This TRAIN might contribute to training them by obtaining phenomenological knowledge of severe accidents, understanding plant vulnerabilities, and solving problems under high stress. (author)

  20. Development and testing of a prototype instrumented bicycle model for the prevention of cyclist accidents

    OpenAIRE

    Miah, S.; Kaparias, I.; Liatsis, P.

    2015-01-01

    Cycling is an increasingly popular mode of travel in cities owing to the great advantages that it offers in terms of space consumption, health and environmental sustainability, and is therefore favoured and promoted by many city authorities worldwide. However, cycling is also perceived as relatively unsafe, and therefore it has yet to be adopted as a viable alternative to the private car. Rising accident numbers, unfortunately, confirm this perception as reality, with a particular source of h...

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

    International Nuclear Information System (INIS)

    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)

  2. Modeling accident frequency in Denmark for improving road safety

    DEFF Research Database (Denmark)

    Lyckegaard, Allan; Hels, Tove; Kaplan, Sigal;

    the infrastructure characteristics and the traffic conditions of the road. The model can be used to point out high risk road segments and support road authorities in planning interventions for the improvement of road safety on Danish roads. The number of accidents on a road link was modeled using a count model after......Traffic accidents result in huge costs to society in terms of death, injury, lost productivity, and property damage. The main objective of the current study is the development of an accident frequency model that predicts the expected number of accidents on a given road segment, provided...... verifying the presence of overdispersion in the variance of the counts. The model relates the number of accidents to the characteristics of the infrastructure in terms of geometry and traffic in order to identify which risk factors and at which degree relate to the possibility of accident occurrence. Data...

  3. 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...... 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 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...

  4. 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

  5. A SCOPING STUDY: Development of Probabilistic Risk Assessment Models for Reactivity Insertion Accidents During Shutdown In U.S. Commercial Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    S. Khericha

    2011-06-01

    This report documents the scoping study of developing generic simplified fuel damage risk models for quantitative analysis from inadvertent reactivity insertion events during shutdown (SD) in light water pressurized and boiling water reactors. In the past, nuclear fuel reactivity accidents have been analyzed both mainly deterministically and probabilistically for at-power and SD operations of nuclear power plants (NPPs). Since then, many NPPs had power up-rates and longer refueling intervals, which resulted in fuel configurations that may potentially respond differently (in an undesirable way) to reactivity accidents. Also, as shown in a recent event, several inadvertent operator actions caused potential nuclear fuel reactivity insertion accident during SD operations. The set inadvertent operator actions are likely to be plant- and operation-state specific and could lead to accident sequences. This study is an outcome of the concern which arose after the inadvertent withdrawal of control rods at Dresden Unit 3 in 2008 due to operator actions in the plant inadvertently three control rods were withdrawn from the reactor without knowledge of the main control room operator. The purpose of this Standardized Plant Analysis Risk (SPAR) Model development project is to develop simplified SPAR Models that can be used by staff analysts to perform risk analyses of operating events and/or conditions occurring during SD operation. These types of accident scenarios are dominated by the operator actions, (e.g., misalignment of valves, failure to follow procedures and errors of commissions). Human error probabilities specific to this model were assessed using the methodology developed for SPAR model human error evaluations. The event trees, fault trees, basic event data and data sources for the model are provided in the report. The end state is defined as the reactor becomes critical. The scoping study includes a brief literature search/review of historical events, developments of

  6. 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

  7. AN IMPROVED MODEL FOR ESTIMATING RUNWAY ACCIDENT COST IN NIGERIA

    Directory of Open Access Journals (Sweden)

    Akinyemi Olasunkanmi Oriola

    2016-03-01

    Full Text Available This research proposes an improved model for estimating runway accident cost in the aviation industry in Nigeria. Bayesian Network was used to model the probability of consequences of runway accidents and subsequently the cost of the potential consequences. The Bayesian Network was also used to implement causal and diagnostic inference. Market interest rate which incorporates the effect of inflation was included to relax the assumption of constant economic value. The three classes of consequences of runway accidents identified in this study were fatal, serious, and minor. Domain experts were used to obtain relevant Bayesian causes and evidences related to the occurrence of consequence of runway accident. A mathematical equation was developed to solve the Bayesian Network influence diagram to obtain the probability of minor runway accident, serious runway accident and fatal runway accident as 0.7603, 0.1547 and 0.0850 respectively. Consequently, the estimated cost of runway accident (minor, serious and fatal was $23, 813.52, $20, 052.64 and $772, 856.06 respectively. The Bayesian Network diagnostic inference reveals the close relation of runway accidents in Nigeria aviation sector with aircraft system failure, approach/takeoff procedures, human factors, weather conditions and collision risk.

  8. Enhanced Accident Tolerant LWR Fuels: Metrics Development

    Energy Technology Data Exchange (ETDEWEB)

    Shannon Bragg-Sitton; Lori Braase; Rose Montgomery; Chris Stanek; Robert Montgomery; Lance Snead; Larry Ott; Mike Billone

    2013-09-01

    The Department of Energy (DOE) Fuel Cycle Research and Development (FCRD) Advanced Fuels Campaign (AFC) is conducting research and development on enhanced Accident Tolerant Fuels (ATF) for light water reactors (LWRs). This mission emphasizes the development of novel fuel and cladding concepts to replace the current zirconium alloy-uranium dioxide (UO2) fuel system. The overall mission of the ATF research is to develop advanced fuels/cladding with improved performance, reliability and safety characteristics during normal operations and accident conditions, while minimizing waste generation. The initial effort will focus on implementation in operating reactors or reactors with design certifications. To initiate the development of quantitative metrics for ATR, a LWR Enhanced Accident Tolerant Fuels Metrics Development Workshop was held in October 2012 in Germantown, MD. This paper summarizes the outcome of that workshop and the current status of metrics development for LWR ATF.

  9. Development of solution behavior observation system under criticality accident conditions

    International Nuclear Information System (INIS)

    A solution behavior observation system was developed for observing the behavior of fissile solution and radiolytic voids under criticality accident conditions in TRACY. The system consisted of a radiation-resistive optical fiberscope and a CCD color video camera. The system functioned properly in the mixed high radiation fields of gamma rays and neutrons under criticality accident conditions, and it succeeded in taking the images of their behavior. They provide an important information to understand phenomena of fuel solution at criticality accidents and to construct computational kinetic models. The images can also be used as teaching materials for plant workers and students in universities. (author)

  10. 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).

  11. 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.

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

    International Nuclear Information System (INIS)

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

  13. 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. PMID:27070081

  14. Development of preliminary Nevada transportation accident characteristics

    International Nuclear Information System (INIS)

    The US DOE, Yucca Mountain Site Characterization Project Office (YMSCPO) has been given the responsibility for characterization of the potential repository site at Yucca Mountain, Nevada, and the analysis of repository-related impacts in the State of Nevada. In support of these responsibilities, the YMSCPO initiated a preliminary study to develop background information on the character of the transportation accidents occurring on the highways and raillines in the State of Nevada. The results of this preliminary study shows that while the transportation accidents in Nevada follow national trends, there are some distinct differences between Nevada and the rest of the Nation. This paper summarizes those results

  15. 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.

  16. 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)

    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

  17. 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.

  18. Mathematical models for steam generator accident simulation

    International Nuclear Information System (INIS)

    In this contribution, the numerical methods used in the DeBeNe-LMFBR development for the analysis of the hydrodynamic and mechanical consequences of steam generator accidents are presented. At first the definition of the source term, i.e. the water leak rate which has to be assumed in the design basis accident as well as the thermochemistry of the sodium/water-reaction is discussed. Then the computer-codes presently used to describe the hydrodynamic and mechanical consequences of steam generator accidents on the basis of the above mentioned source term are presented. These comprise the code-system SAPHYR and the code PTANER and PISCES. Furthermore, developments which are planned or already under way for future use, such as the BEREPOT-code, are presented. (author)

  19. A MELCOR model of Fukushima Daiichi Unit 3 accident

    International Nuclear Information System (INIS)

    Highlights: • A MELCOR model of the Fukushima Unit 3 accident was developed. • The MELCOR input file is published as electronic supplementary data with this paper. • Reactor pressure vessel lower head failed about 53 h after the earthquake. • 70% of fuel was discharged from reactor to containment. • 0.95% of cesium inventory was released to the environment. - Abstract: A MELCOR model of the Fukushima Daiichi Unit 3 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as electronic supplementary data with this paper. According to the calculation, the reactor pressure vessel lower head failed about 53 h after the earthquake. At the end of the calculation, 30% of the fuel was still inside the reactor and 70% had been discharged to the containment. Almost all of the radioactive noble gases and 0.95% of the cesium inventory were released to the environment during the accident

  20. A MELCOR model of Fukushima Daiichi Unit 3 accident

    Energy Technology Data Exchange (ETDEWEB)

    Sevón, Tuomo, E-mail: tuomo.sevon@vtt.fi

    2015-04-01

    Highlights: • A MELCOR model of the Fukushima Unit 3 accident was developed. • The MELCOR input file is published as electronic supplementary data with this paper. • Reactor pressure vessel lower head failed about 53 h after the earthquake. • 70% of fuel was discharged from reactor to containment. • 0.95% of cesium inventory was released to the environment. - Abstract: A MELCOR model of the Fukushima Daiichi Unit 3 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as electronic supplementary data with this paper. According to the calculation, the reactor pressure vessel lower head failed about 53 h after the earthquake. At the end of the calculation, 30% of the fuel was still inside the reactor and 70% had been discharged to the containment. Almost all of the radioactive noble gases and 0.95% of the cesium inventory were released to the environment during the accident.

  1. Development and test results of the Realtime Severe Accident Model 5 (RSAM5) based on the MAAP5 For the Kori 1 simulator

    Energy Technology Data Exchange (ETDEWEB)

    Hong, Jin Hyuk; Lee, Myeong Soo [KHNP Central Research Institute, Daejeon (Korea, Republic of)

    2012-10-15

    The Real Time Severe Accident Model (RSAM) in the Kori simulator employs the standard MAAP 5.01.1101 code (which is defined as MAAP 5.01) plus several statically linked libraries that interface with the simulator environment. The physical phenomena that can be envisioned inside the reactor vessel, the reactor coolant system (RCS), and the containment during severe accidents are comprehensively modeled by the MAAP5 code. The MAAP5 code has been known to be a reliable tool for understanding the sequence of events that occur during severe LWR accidents, evaluating the consequences of the failure of emergency systems, assessing the effects of operator interventions, and investigating the influence of design features of the RCS, containment, and safety systems on the accident consequences. The purpose of this paper is to describe the modeling of the Kori Unit 1 nuclear plant with the MAAP5 code and major outputs in the event of the SBO, SBO + SGTR, SBO + LBLOCA.

  2. Modeling alternative clad behavior for accident tolerant systems

    International Nuclear Information System (INIS)

    The US Department of Energy Fuel Cycle Research and Development program has a key goal of helping develop accident tolerant fuels (ATF) through investigating fuel and clad forms. In the current work thermochemical modeling and experiment are being used to assess fuel and clad alternatives. Cladding alternatives that have promise to improve fuel performance under accident conditions include the FeCrAl family of alloys and SiC-based composites. These are high strength and radiation resistant alloys and ceramics that have increased resistance to oxidation as compared to zirconium alloys. Accident modeling codes have indicated substantially increased time to failure and resulting effects. In the current work the thermochemical behavior of these materials are being assessed and the work reported here. (author)

  3. 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. PMID:23376544

  4. 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.

  5. Accident evolution and barrier function and accident evolution management modeling of nuclear power plant incidents

    International Nuclear Information System (INIS)

    Every analysis of an accident or an incident is founded on a more or less explicit model of what an accident is. On a general level, the current approach models an incident or accident in a nuclear power plant as a failure to maintain a stable state with all variables within their ranges of stability. There are two main sets of subsystems in continuous interaction making up the analyzed system, namely the human-organizational and the technical subsystems. Several different but related approaches can be chosen to model an accident. However, two important difficulties accompany such modeling: the high level of system complexity and the very infrequent occurrence of accidents. The current approach acknowledges these problems and focuses on modeling reported incidents/accidents or scenarios selected in probabilistic risk assessment analyses to be of critical importance for the safety of a plant

  6. Developing techniques for cause-responsibility analysis of occupational accidents.

    Science.gov (United States)

    Jabbari, Mousa; Ghorbani, Roghayeh

    2016-11-01

    The aim of this study was to specify the causes of occupational accidents, determine social responsibility and the role of groups involved in work-related accidents. This study develops occupational accidents causes tree, occupational accidents responsibility tree, and occupational accidents component-responsibility analysis worksheet; based on these methods, it develops cause-responsibility analysis (CRA) techniques, and for testing them, analyzes 100 fatal/disabling occupational accidents in the construction setting that were randomly selected from all the work-related accidents in Tehran, Iran, over a 5-year period (2010-2014). The main result of this study involves two techniques for CRA: occupational accidents tree analysis (OATA) and occupational accidents components analysis (OACA), used in parallel for determination of responsible groups and responsibilities rate. From the results, we find that the management group of construction projects has 74.65% responsibility of work-related accidents. The developed techniques are purposeful for occupational accidents investigation/analysis, especially for the determination of detailed list of tasks, responsibilities, and their rates. Therefore, it is useful for preventing work-related accidents by focusing on the responsible group's duties.

  7. Validation and verification of accident consequence assessment models

    Energy Technology Data Exchange (ETDEWEB)

    Homma, T.; Togawa, O. [Japan Atomic Energy Research Inst., Tokyo (Japan); Takahashi, T. [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Arkhipov, A.N. [Chernobyl Science and Technology Centre for International Research (Ukraine)

    2001-03-01

    An accident consequence assessment code, OSCAAR, primarily designed by Japan Atomic Energy Research Institute (JAERI) for use in probabilistic safety assessment (PSA) of nuclear reactors in Japan, was applied to use for siting, emergency planning, and development of design criteria, and in the comparative risk studies of different energy systems. After verifying the code system through the international code comparison organized by CEC and OECD/NEA, the validation and improvements of the individual models and the verification of the whole OSCAAR code system were made. The cooperative research between Chernobyl Science and Technology Center for International Research (CHESCIR) and JAERI provided a valuable opportunity to test the performance of the accident consequence assessment models by comparing the model predictions with data obtained in the Chernobyl accidents. The predictive capabilities of OSCAAR were demonstrated using the accident source term and meteorological data for estimating the early exposure to the public occurred during and shortly after plume passage. The calculations indicated that ground-shine dose and inhalation dose, particularly from large nonvolatile particulates were the main contributors in the early stage of the accident. (S. Ohno)

  8. Advanced accident sequence precursor analysis level 2 models

    Energy Technology Data Exchange (ETDEWEB)

    Galyean, W.J.; Brownson, D.A.; Rempe, J.L. [and others

    1996-03-01

    The U.S. Nuclear Regulatory Commission Accident Sequence Precursor program pursues the ultimate objective of performing risk significant evaluations on operational events (precursors) occurring in commercial nuclear power plants. To achieve this objective, the Office of Nuclear Regulatory Research is supporting the development of simple probabilistic risk assessment models for all commercial nuclear power plants (NPP) in the U.S. Presently, only simple Level 1 plant models have been developed which estimate core damage frequencies. In order to provide a true risk perspective, the consequences associated with postulated core damage accidents also need to be considered. With the objective of performing risk evaluations in an integrated and consistent manner, a linked event tree approach which propagates the front end results to back end was developed. This approach utilizes simple plant models that analyze the response of the NPP containment structure in the context of a core damage accident, estimate the magnitude and timing of a radioactive release to the environment, and calculate the consequences for a given release. Detailed models and results from previous studies, such as the NUREG-1150 study, are used to quantify these simple models. These simple models are then linked to the existing Level 1 models, and are evaluated using the SAPHIRE code. To demonstrate the approach, prototypic models have been developed for a boiling water reactor, Peach Bottom, and a pressurized water reactor, Zion.

  9. 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.

  10. A model national emergency plan for radiological accidents

    International Nuclear Information System (INIS)

    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

  11. Catastrophe model of the accident process, safety climate, and anxiety.

    Science.gov (United States)

    Guastello, Stephen J; Lynn, Mark

    2014-04-01

    This study aimed (a) to address the evidence for situational specificity in the connection between safety climate to occupational accidents, (b) to resolve similar issues between anxiety and accidents, (c) to expand and develop the concept of safety climate to include a wider range of organizational constructs, (d) to assess a cusp catastrophe model for occupational accidents where safety climate and anxiety are treated as bifurcation variables, and environ-mental hazards are asymmetry variables. Bifurcation, or trigger variables can have a positive or negative effect on outcomes, depending on the levels of asymmetry, or background variables. The participants were 1262 production employees of two steel manufacturing facilities who completed a survey that measured safety management, anxiety, subjective danger, dysregulation, stressors and hazards. Nonlinear regression analyses showed, for this industry, that the accident process was explained by a cusp catastrophe model in which safety management and anxiety were bifurcation variables, and hazards, age and experience were asymmetry variables. The accuracy of the cusp model (R2 = .72) exceeded that of the next best log-linear model (R2 = .08) composed from the same survey variables. The results are thought to generalize to any industry where serious injuries could occur, although situationally specific effects should be anticipated as well.

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

    International Nuclear Information System (INIS)

    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

  13. Development of severe accident management and training support system

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-04-01

    Recently, the overall severe accident management strategy is under development according to the logical flow of severe accident management guidelines in some foreign countries. In Korea, the basis of severe accident management strategy is established due to the development of Korean severe accident guideline. In the straining system, the professional information as well as the general information for severe accident should be provided to the related personnel and the function of prior simulation for plant behavior according to strategy execution should be required. Korean severe accident management guideline is chosen as the basis logic for development of support system for decision-support and training related with execution of severe accident strategy. The training simulator is developed for prior expectation of plant behavior and the severe accident computer code, MELCOR, is utilized as the engine, and it is possible to operate equipments necessary for execution of severe accident management guidelines. And also, the graphical interface is developed to provide the plant status and provide status change of major equipments dynamically.

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

    International Nuclear Information System (INIS)

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

  16. 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.

  17. MELCOR modeling of Fukushima unit 2 accident

    International Nuclear Information System (INIS)

    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.

  18. 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.

  19. Development of Database for Accident Analysis in Indian Mines

    Science.gov (United States)

    Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

    2015-08-01

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

  20. Health effects estimation code development for accident consequence analysis

    International Nuclear Information System (INIS)

    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)

  1. Development of the Severe Accident Analysis DB for the Severe Accident Management Expert System (I)

    Energy Technology Data Exchange (ETDEWEB)

    Park, Soo Yong; Ahn, Kwang Il [KAERI, Daejeon (Korea, Republic of)

    2010-12-15

    This report contains analysis methodologies and calculation results of 5 initiating events of the severe accident analysis database system. The Ulchin 3,4 NPP has been selected as reference plants. Based on the probabilistic safety analysis of the corresponding plant, 54 accident scenarios, which was predicted to have more than 10-10 /ry occurrence frequency, have been analyzed as base cases for the Large loss of Coolant sequence database. The functions of the severe accident analysis database system will be to make a diagnosis of the accident by some input information from the plant symptoms, to search a corresponding scenario, and finally to provide the user phenomenological information based on the pre-analyzed results. The MAAP 4.06 calculation results in this report will be utilized as input data to develop the database system

  2. Applying speed prediction models to define road sections and to develop accident prediction models : A German case study and a Portuguese exploratory study. Road Infrastructure Safety Management Evaluation Tools (RISMET), Deliverable No. 6.2.

    NARCIS (Netherlands)

    Dietze, M. & Weller, G.

    2014-01-01

    Traditional concepts of accident prediction models and safety performance functions are based on certain input values derived from infrastructure and accident occurrence. Usually, driving behaviour is not taken into account even though its impact is known. So far, an implementation of driving behavi

  3. Lessons Learned from the Past Accidents for Safety Culture Development

    International Nuclear Information System (INIS)

    All nuclear organizations strive to sustain and improve safety. There is diversity in the way organizations understand the concept of safety and the actions that can help to drive improvements. This paper presents an overview of the lessons to be learned from past nuclear accidents and their relevance for the development of nuclear safety culture. Although the term Safety Culture emerged after the Chernobyl accident, the factors that contributed to earlier accidents, of which the most notable was the accident of Three Mile Island Unit 2 , are also relevant for nuclear safety culture. As regards the Fukushima accident from 2011, safety culture was once again brought into discussion. It is easier to manage the workplaces and the organizations than the minds of employees, as it is not possible to change the human condition, but changing the conditions under which people work. For this, the commitment of the top management is important, without which, it is not possible to make the necessary changes. (author)

  4. Development of a mathematical model for studying rewetting of reactor fuel elements after a loss-of-coolant accident

    International Nuclear Information System (INIS)

    The ZETHYF model allows studying the flow, heat transmission and temperature conditions in a cooling channel and, thus, recretting of reactor fuel elements. Coolant flow is calculated for a single-phase or dual-phase coolant along coolant channel on the assumption of constant pressure. Within dual-phase flow, a thermodynamically balanced slip flow is assumed for small steam volume proportions and saturation temperature is assumed for water still existing in the form of droplets. Droplet velocity is represented by means of a momentum balance between steam and droplet, evaporation of droplets is established by means of heat supply from steam, cladding tube wall and by thermal radiation. Minimum film boiling temperature is taken as a criterium for secretting and is compared with local cladding tube temperature which is established by way of a 2-D thermal conduction equation check calculation of experiments (Flecht) yielded sufficient agreement. Calculation results yielded a somewhat slower cooling of heating element and, consequently, delayed rewetting. (orig./HP)

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

  6. The accident evolution and barrier model applied to incident analysis in the processing industries

    International Nuclear Information System (INIS)

    This study presents a model for how accidents develop and how the accident evolution can be arrested. The model describes the interaction between the technical and human-organizational systems which may lead to an accident. The framework provided by the model may be used in predictive safety analyses as well as in post-hoc incident analyses. To illustrate this, the model is applied on an incident reported by the nuclear industry of Sweden. In general, application of the model will indicate where safety can be improved and raises questions about issues such as the cost, feasibility and effectiveness of different ways of increasing safety. (author). 15 refs, 2 figs

  7. Advanced accident sequence precursor analysis level 1 models

    Energy Technology Data Exchange (ETDEWEB)

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K.; Schroeder, J.A.; Siu, N.O. [Idaho National Engineering Lab., Idaho National Lab., Idaho Falls, ID (United States)

    1996-03-01

    INEL has been involved in the development of plant-specific Accident Sequence Precursor (ASP) models for the past two years. These models were developed for use with the SAPHIRE suite of PRA computer codes. They contained event tree/linked fault tree Level 1 risk models for the following initiating events: general transient, loss-of-offsite-power, steam generator tube rupture, small loss-of-coolant-accident, and anticipated transient without scram. Early in 1995 the ASP models were revised based on review comments from the NRC and an independent peer review. These models were released as Revision 1. The Office of Nuclear Regulatory Research has sponsored several projects at the INEL this fiscal year to further enhance the capabilities of the ASP models. Revision 2 models incorporates more detailed plant information into the models concerning plant response to station blackout conditions, information on battery life, and other unique features gleaned from an Office of Nuclear Reactor Regulation quick review of the Individual Plant Examination submittals. These models are currently being delivered to the NRC as they are completed. A related project is a feasibility study and model development of low power/shutdown (LP/SD) and external event extensions to the ASP models. This project will establish criteria for selection of LP/SD and external initiator operational events for analysis within the ASP program. Prototype models for each pertinent initiating event (loss of shutdown cooling, loss of inventory control, fire, flood, seismic, etc.) will be developed. A third project concerns development of enhancements to SAPHIRE. In relation to the ASP program, a new SAPHIRE module, GEM, was developed as a specific user interface for performing ASP evaluations. This module greatly simplifies the analysis process for determining the conditional core damage probability for a given combination of initiating events and equipment failures or degradations.

  8. Grey-Markov Model for Road Accidents Forecasting

    Institute of Scientific and Technical Information of China (English)

    李相勇; 严余松; 蒋葛夫

    2003-01-01

    In order to improve the forecasting precision of road accidents, by introducing Markov chains forecasting method, a grey-Markov model for forecasting road accidents is established based on grey forecasting method. The model combines the advantages of both grey forecasting method and Markov chains forecasting method, overcomes the influence of random fluctuation data on forecasting precision and widens the application scope of the grey forecasting. An application example is conducted to evaluate the grey-Markov model, which shows that the precision of the grey-Markov model is better than that of grey model in forecasting road accidents.

  9. Thermal-hydraulic modeling of reactivity accidents in MTR reactors

    Directory of Open Access Journals (Sweden)

    Khater Hany

    2006-01-01

    Full Text Available This paper describes the development of a dynamic model for the thermal-hydraulic analysis of MTR research reactors during a reactivity insertion accident. The model is formulated for coupling reactor kinetics with feedback reactivity and reactor core thermal-hydraulics. To represent the reactor core, two types of channels are considered, average and hot channels. The developed computer program is compiled and executed on a personal computer, using the FORTRAN language. The model is validated by safety-related benchmark calculations for MTR-TYPE reactors of IAEA 10 MW generic reactor for both slow and fast reactivity insertion transients. A good agreement is shown between the present model and the benchmark calculations. Then, the model is used for simulating the uncontrolled withdrawal of a control rod of an ETRR-2 reactor in transient with over power scram trip. The model results for ETRR-2 are analyzed and discussed.

  10. 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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    Jang, S. H.; Chun, S. W.; Jang, H. S. and others [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

    1993-01-15

    As a continuing study for the development of a system of computer codes to analyze severe accidents which had been performed last year, major focuses were on the aspect of application of the developed code systems. As the first step, two most commonly used code packages other than STCP, i.e., MELCOR of NRC and MAAP of IDCOR were reviewed to compare the models that they used. Next, important heat transfer phenomena were surveyed as severe accident progressed. Particularly, debris bed coolability and molten core-concrete interaction were selected as sample models, and they were studied extensively. The recent theoretical works and experiments for these phenomena were surveyed, and also the relevant models adopted by major code packages were compared and assessed. Based on the results obtained in this study, it is expected to be able to take into account these phenomenological uncertainties when one uses the severe accident code packages for probabilistic safety assessments or accident management programs.

  13. 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.

  14. A MELCOR model of Fukushima Daiichi Unit 1 accident

    International Nuclear Information System (INIS)

    Highlights: • A MELCOR model of Fukushima Unit 1 accident was developed. • The MELCOR input file is published as Electronic Supplementary data with this paper. • Molten fuel was discharged to containment from broken reactor pressure vessel. • Almost all radioactive noble gases and about 0.05% of cesium inventory were released to the environment. • Calculated release rates from Units 1, 2, and 3 were compared with measured radiation dose rate. - Abstract: A MELCOR model of Fukushima Daiichi Unit 1 accident was developed. The model is based on publicly available information, and the MELCOR input file is published as Electronic Supplementary data with this paper. In order to reproduce the measured containment pressure, it was necessary to model a leak from the reactor coolant system. Recirculation pump seal leak, starting 5 h after the earthquake, was assumed in this study. The reactor pressure vessel lower head was calculated to fail, and all fuel was discharged to the containment. Almost all of the radioactive noble gases and about 0.05% of the cesium inventory were released to the environment, according to this calculation. Calculated release rates from Units 1, 2, and 3 were compared with measured radiation dose rate in the plant area

  15. 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)

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

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    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...

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

    DEFF Research Database (Denmark)

    Lind, Morten; Zhang, Xinxin

    2014-01-01

    The paper investigates 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...

  18. Multiscale Multiphysics Developments for Accident Tolerant Fuel Concepts

    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); Yu, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Zhang, Y. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bai, X. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Andersson, D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Patra, A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Wen, W. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Tome, C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Baskes, M. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Martinez, E. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Stanek, C. R. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Miao, Y. [Argonne National Lab. (ANL), Argonne, IL (United States); Ye, B. [Argonne National Lab. (ANL), Argonne, IL (United States); Hofman, G. L. [Argonne National Lab. (ANL), Argonne, IL (United States); Yacout, A. M. [Argonne National Lab. (ANL), Argonne, IL (United States); Liu, W. [ANATECH Corp., San Diego, CA (United States)

    2015-09-01

    U3Si2 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 U3Si2 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.

  19. Model based detection and reconstruction of road traffic accidents

    OpenAIRE

    Hiemer, Marcus

    2005-01-01

    This thesis describes the detection and reconstruction of traffic accidents with event data recorders. The underlying idea is to describe the vehicle motion and dynamics up to the stability limit by means of linear and non-linear vehicle models. These models are used to categorize the driving behavior and to freeze the recorded data in a memory if an accident occurs. Based on these data, among others the vehicle trajectory is reconstructed with fuzzy data fusion. The side slip angle whi...

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

  2. Risk forecasting and evaluating model of Environmental pollution accident

    Institute of Scientific and Technical Information of China (English)

    ZENG Wei-hua; CHENG Sheng-tong

    2005-01-01

    Environmental risk (ER) fact ore come from ER source and they are controlled by the primary control mechanism (PCM) of environmental risk, due to the self failures or the effects of external environment risk trigger mechanism, the PCM could not work regularly any more, then, the ER factore will release environmental space, and an ER field is formed up. The forming of ER field does not mean that any environmental pollution accident(EPA) will break out; only the ER receptore are exposed in the ER field and damaged seriously,the potential ER really turns into an actual EPA. Researching on the general laws of evolving from environmental risk to EPA, this paper bring forwards a relevant concept model of risk forecasting and evaluating of EPA. This model provides some scientific methods for risk evaluation, prevention and emergency response of EPA. This model not only enriches and develops the theory system of environment safety and emergency response, but also acts as an instruction for public safety, enterprise' s safety management and emergency response of the accident.

  3. Traffic Accident, System Model and Cluster Analysis in GIS

    Directory of Open Access Journals (Sweden)

    Veronika Vlčková

    2015-07-01

    Full Text Available One of the many often frequented topics as normal journalism, so the professional public, is the problem of traffic accidents. This article illustrates the orientation of considerations to a less known context of accidents, with the help of constructive systems theory and its methods, cluster analysis and geoinformation engineering. Traffic accident is reframing the space-time, and therefore it can be to study with tools of technology of geographic information systems. The application of system approach enabling the formulation of the system model, grabbed by tools of geoinformation engineering and multicriterial and cluster analysis.

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

    DEFF Research Database (Denmark)

    Janstrup, Kira Hyldekær

    management tool.Initially models were built by using existing traffic accident data collected by the police and emergency rooms in Denmark. The data registered by the police was collected on traffic accidents occurred on Danish roads in the period between 2002 and 2008. The emergency room data were collected...... to be the most relevant factor related to the lack of intention to report future cycling accidents. Secondly, the factors: concerns about family distress and social image and preference to allocate time to other activities are both associated with non-reporting intentions (Paper 3). 5) New information about...

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

    International Nuclear Information System (INIS)

    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

  6. A web-based nuclear accident illumination system based on multilevel flow model - for risk communication and nuclear safety culture

    International Nuclear Information System (INIS)

    This paper introduces a new method to illuminate the nuclear accident by Multilevel Flow Model, and based on the method, a web-based nuclear accident illumination system is proposed to represent the current nuclear accident in nuclear power plant of Japan in an understandable way. The MFM is a means-end and part-whole modeling method to describe the structure and the intention of a plant process. The relationship between the MFM functions enables accident prediction for a plant process. Thus, a web-based accident illumination system based by MFM can describe the nuclear accident in the nuclear power plant clearly and be accessed by public to make the public get to know and understand the nuclear power and nuclear risk. The public can build their own confidence of the nuclear power by their understanding of the nuclear accident with this system and this is helpful to build a harmonious development environment for nuclear power. (author)

  7. Emergency control system based on the analytical hierarchy process and coordinated development degree model for sudden water pollution accidents in the Middle Route of the South-to-North Water Transfer Project in China.

    Science.gov (United States)

    Long, Yan; Xu, Guobin; Ma, Chao; Chen, Liang

    2016-06-01

    Water transfer projects are important for realizing reasonable allocation of water resources, but once a water pollution accident occurs during such a project, the water environment is exposed to enormous risks. Therefore, it is critical to determine an appropriate emergency control system (ECS) for sudden water pollution accidents that occur in water transfer projects. In this study, the analytical hierarchy process (AHP) integrated with the coordinated development degree model (CDDM) was used to develop the ECS. This ECS was developed into two parts, including the emergency risk assessment and the emergency control. Feasible emergency control targets and control technology were also proposed for different sudden water pollution accidents. A demonstrative project was conducted in the Fangshui to Puyang channel, which is part of the Beijing-Shijiazhuang Emergency Water Supply Project (BSP) in the Middle Route of the South-to-North Water Transfer Project (MR-SNWTP) in China. However, we could not use an actual toxic soluble pollutant to validate our ECS, so we performed the experiment with sucrose to test the ECS based on its concentration variation. The relative error of peak sucrose concentration was less than 20 %. PMID:26979314

  8. Emergency control system based on the analytical hierarchy process and coordinated development degree model for sudden water pollution accidents in the Middle Route of the South-to-North Water Transfer Project in China.

    Science.gov (United States)

    Long, Yan; Xu, Guobin; Ma, Chao; Chen, Liang

    2016-06-01

    Water transfer projects are important for realizing reasonable allocation of water resources, but once a water pollution accident occurs during such a project, the water environment is exposed to enormous risks. Therefore, it is critical to determine an appropriate emergency control system (ECS) for sudden water pollution accidents that occur in water transfer projects. In this study, the analytical hierarchy process (AHP) integrated with the coordinated development degree model (CDDM) was used to develop the ECS. This ECS was developed into two parts, including the emergency risk assessment and the emergency control. Feasible emergency control targets and control technology were also proposed for different sudden water pollution accidents. A demonstrative project was conducted in the Fangshui to Puyang channel, which is part of the Beijing-Shijiazhuang Emergency Water Supply Project (BSP) in the Middle Route of the South-to-North Water Transfer Project (MR-SNWTP) in China. However, we could not use an actual toxic soluble pollutant to validate our ECS, so we performed the experiment with sucrose to test the ECS based on its concentration variation. The relative error of peak sucrose concentration was less than 20 %.

  9. Traffic Accident Prediction Model Implementation in Traffic Safety Management

    OpenAIRE

    Wen, Keyao

    2009-01-01

    As one of the highest fatalities causes, traffic accidents and collisions always requires a large amounteffort to be reduced or prevented from occur. Traffic safety management routines therefore always needefficient and effective implementation due to the variations of traffic, especially from trafficengineering point of view apart from driver education.Traffic Accident Prediction Model, considered as one of the handy tool of traffic safety management,has become of well followed with interest...

  10. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  13. Development of emergency response support system for accident management

    International Nuclear Information System (INIS)

    Specific measures for the accident management (AM) are proposed to prevent the severe accident and to mitigate their effects in order to upgrade the safety of nuclear power plants even further. To ensure accident management effective, it is essential to grasp the plant status accurately. In consideration of the above mentioned background, the Emergency Response Support System (ERSS) was developed as a computer assisted prototype system by a joint study of Japanese BWR group. This system judges and predicts the plant status at the emergency condition in a nuclear power plant. This system displays the results of judgment and prediction. The effectiveness of the system was verified through the test and good prospects for applying the system to a plant was obtained. 7 refs., 10 figs

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

    OpenAIRE

    Martin, Robert P.

    2012-01-01

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

  15. Estimation of traffic accident costs: a prompted model.

    Science.gov (United States)

    Hejazi, Rokhshad; Shamsudin, Mad Nasir; Radam, Alias; Rahim, Khalid Abdul; Ibrahim, Zelina Zaitun; Yazdani, Saeed

    2013-01-01

    Traffic accidents are the reason for 25% of unnatural deaths in Iran. The main objective of this study is to find a simple model for the estimation of economic costs especially in Islamic countries (like Iran) in a straightforward manner. The model can show the magnitude of traffic accident costs with monetary equivalent. Data were collected from different sources that included traffic police records, insurance companies and hospitals. The conceptual framework, in our study, was based on the method of Ayati. He used this method for the estimation of economic costs in Iran. We promoted his method via minimum variables. Our final model has only three available variables which can be taken from insurance companies and police records. The running model showed that the traffic accident costs were US$2.2 million in 2007 for our case study route.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1995-07-01

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

  17. 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

  18. 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.

  19. Quantifying safety benefit of winter road maintenance: accident frequency modeling.

    Science.gov (United States)

    Usman, Taimur; Fu, Liping; Miranda-Moreno, Luis F

    2010-11-01

    This research presents a modeling approach to investigate the association of the accident frequency during a snow storm event with road surface conditions, visibility and other influencing factors controlling for traffic exposure. The results have the premise to be applied for evaluating different maintenance strategies using safety as a performance measure. As part of this approach, this research introduces a road surface condition index as a surrogate measure of the commonly used friction measure to capture different road surface conditions. Data from various data sources, such as weather, road condition observations, traffic counts and accidents, are integrated and used to test three event-based models including the Negative Binomial model, the generalized NB model and the zero inflated NB model. These models are compared for their capability to explain differences in accident frequencies between individual snow storms. It was found that the generalized NB model best fits the data, and is most capable of capturing heterogeneity other than excess zeros. Among the main results, it was found that the road surface condition index was statistically significant influencing the accident occurrence. This research is the first showing the empirical relationship between safety and road surface conditions at a disaggregate level (event-based), making it feasible to quantify the safety benefits of alternative maintenance goals and methods.

  20. 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.

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

    International Nuclear Information System (INIS)

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

  2. Modeling and assessment of accident consequences; development of RODOS, a real-time on-line decision support system for nuclear emergencies in Europe

    International Nuclear Information System (INIS)

    In cooperation with NRPB (UK), the first version 1.0 of PC COSYMA for use on advanced PCs has been released; during a training course in mid 1993, future users were educated in operating the software. The main frame version of the program package COSYMA has been up-dated with new dose conversion factors and fodd-chain data and was distributed to some 20 institutes in Europe and abroad. The comparative calculations performed within the international OECD(NEA)/CEC intercomparison of accident consequence assessment codes were analysed and documented in three reports. Furtheron, consequence assessments have been performed for the research reactor BER II (two source terms) and documented; the influence on individual doses and emergency actions of inplant accident management measures in future EPRs was quantified; within th scope of a EC/US-study on the external costs of the energy cycle, accident consequences were assessed for three source terms. (orig.)

  3. 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)

  4. 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.

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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.)

  7. Final safety analysis report for the Galileo Mission: Volume 2, Book 2: Accident model document: Appendices

    Energy Technology Data Exchange (ETDEWEB)

    1988-12-15

    This section of the Accident Model Document (AMD) presents the appendices which describe the various analyses that have been conducted for use in the Galileo Final Safety Analysis Report II, Volume II. Included in these appendices are the approaches, techniques, conditions and assumptions used in the development of the analytical models plus the detailed results of the analyses. Also included in these appendices are summaries of the accidents and their associated probabilities and environment models taken from the Shuttle Data Book (NSTS-08116), plus summaries of the several segments of the recent GPHS safety test program. The information presented in these appendices is used in Section 3.0 of the AMD to develop the Failure/Abort Sequence Trees (FASTs) and to determine the fuel releases (source terms) resulting from the potential Space Shuttle/IUS accidents throughout the missions.

  8. Development status of Severe Accident Analysis Code SAMPSON

    Energy Technology Data Exchange (ETDEWEB)

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

    2000-11-01

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

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

    International Nuclear Information System (INIS)

    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

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

    OpenAIRE

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

    2010-01-01

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

  12. Modelling and analysis of the behavior of LWRs at severe core accidents

    International Nuclear Information System (INIS)

    With respect to the assessment of the consequences of severe accidents in light water reactors from the initiation of the accident up to the thermal failure of the reactor pressure vessel (RPV), a modular program system has been developed. Experimental results will be considered with respect to the modeling of the fuel rod behavior, e.g. deformation of the fuel rod, metal water reaction and the melting of the fuel rods. The fuel and core models allow to estimate the coolability of fuel rods and core as well as the consequences of core meltdown accidents at various pressure levels. After partial failure of the lower core retention structure, the core material will drop into the lower plenum and heat up the RPV. This strong interaction between the thermal behavior of the remaining core and the partially dropped core material has been modeled because of an accident sequence analysis. The analyses described here show, that not the entire core will fail, but a partial drop of core material into the lower plenum is likely to occur. With respect to the validation of the program system, comparison calculations with the fuel rod behavior and melt models SSYST and EXMEL will be performed. Moreover, the program system will be applied to the bundle behavior in meltdown experiments, the TMI-2 core behavior and the course of a core meltdown accident in risk studies. (orig.)

  13. 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.

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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

  16. 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

  17. Accident prediction model for public highway-rail grade crossings.

    Science.gov (United States)

    Lu, Pan; Tolliver, Denver

    2016-05-01

    Considerable research has focused on roadway accident frequency analysis, but relatively little research has examined safety evaluation at highway-rail grade crossings. Highway-rail grade crossings are critical spatial locations of utmost importance for transportation safety because traffic crashes at highway-rail grade crossings are often catastrophic with serious consequences. The Poisson regression model has been employed to analyze vehicle accident frequency as a good starting point for many years. The most commonly applied variations of Poisson including negative binomial, and zero-inflated Poisson. These models are used to deal with common crash data issues such as over-dispersion (sample variance is larger than the sample mean) and preponderance of zeros (low sample mean and small sample size). On rare occasions traffic crash data have been shown to be under-dispersed (sample variance is smaller than the sample mean) and traditional distributions such as Poisson or negative binomial cannot handle under-dispersion well. The objective of this study is to investigate and compare various alternate highway-rail grade crossing accident frequency models that can handle the under-dispersion issue. The contributions of the paper are two-fold: (1) application of probability models to deal with under-dispersion issues and (2) obtain insights regarding to vehicle crashes at public highway-rail grade crossings. PMID:26922288

  18. Modelling of conspicuity-related motorcycle accidents in Seremban and Shah Alam, Malaysia.

    Science.gov (United States)

    Radin, U R; Mackay, M G; Hills, B L

    1996-05-01

    Preliminary analysis of the short-term impact of a running headlights intervention revealed that there has been a significant drop in conspicuity-related motorcycle accidents in the pilot areas, Seremban and Shah Alam, Malaysia. This paper attempts to look in more detail at conspicuity-related accidents involving motorcycles. The aim of the analysis was to establish a statistical model to describe the relationship between the frequency of conspicuity-related motorcycle accidents and a range of explanatory variables so that new insights can be obtained into the effects of introducing a running headlight campaign and regulation. The exogenous variables in this analysis include the influence of time trends, changes in the recording and analysis system, the effect of fasting activities during Ramadhan and the "Balik Kampong" culture, a seasonal cultural-religious holiday activity unique to Malaysia. The model developed revealed that the running headlight intervention reduced the conspicuity-related motorcycle accidents by about 29%. It is concluded that the intervention has been successful in improving conspicuity-related motorcycle accidents in Malaysia. PMID:8799436

  19. Accident consequence assessments with different atmospheric dispersion models

    International Nuclear Information System (INIS)

    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.)

  20. Modelling the oil spill track from Prestige-Nassau accident

    Science.gov (United States)

    Montero, P.; Leitao, P.; Penabad, E.; Balseiro, C. F.; Carracedo, P.; Braunschweig, F.; Fernandes, R.; Gomez, B.; Perez-Munuzuri, V.; Neves, R.

    2003-04-01

    On November 13th 2002, the tank ship Prestige-Nassau sent a SOS signal. The hull of the ship was damaged producing an oil spill in front of the Galician coast (NW Spain). The damaged ship took north direction spilling more fuel and affecting the western Galician coast. After this, it changed its track to south. At this first stage of the accident, the ship spilt around 10000 Tm in 19th at the Galician Bank, at 133 NM of Galician coast. From the very beginning, a monitoring and forecasting of the first slick was developed. Afterwards, since southwesternly winds are frequent in wintertime, the slick from the initial spill started to move towards the Galician coast. This drift movement was followed by overflights. With the aim of forecasting the place and arriving date to the coast, some simulations with two different models were developed. The first one was a very simple drift model forced with the surface winds generated by ARPS operational model (1) at MeteoGalicia (regional weather forecast service). The second one was a more complex hydrodynamic model, MOHID2000 (2,3), developed by MARETEC GROUP (Instituto Superior Técnico de Lisboa) in collaboration with GFNL (Grupo de Física Non Lineal, Universidade de Santiago de Compostela). On November 28th, some tarballs appeared at south of main slick. This observations could be explained taking into account the below surface water movement following Ekman dynamic. Some new simulations with the aim of understanding better the physic underlying these observations were performed. Agreed between observations and simulations was achieved. We performed simulations with and without slope current previously calculated by other authors, showing that this current can only introduce subtle differences in the slick's arriving point to the coast and introducing wind as the primary forcing. (1) A two-dimensional particle tracking model for pollution dispersion in A Coruña and Vigo Rias (NW Spain). M. Gómez-Gesteira, P. Montero, R

  1. 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

  2. Multilevel modelling for the regional effect of enforcement on road accidents.

    Science.gov (United States)

    Yannis, George; Papadimitriou, Eleonora; Antoniou, Constantinos

    2007-07-01

    This paper investigates the effect of the intensification of Police enforcement on the number of road accidents at national and regional level in Greece, focusing on one of the most important road safety violations: drinking-and-driving. Multilevel negative binomial models are developed to describe the effect of the intensification of alcohol enforcement on the reduction of road accidents in different regions of Greece. Moreover, two approaches are explored as far as regional clustering is concerned: the first one concerns an ad hoc geographical clustering and the second one is based on the results of mathematical cluster analysis through demographic, transport and road safety characteristics. Results indicate that there are significant spatial dependences among road accidents and enforcement. Additionally, it is shown that these dependences are more efficiently interpreted when regions are determined on the basis of qualitative similarities than on the basis of geographical adjacency. PMID:17274938

  3. The development of a nuclear accident risk information system

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. [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. PMID:15073638

  7. 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.

  8. 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)

    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

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

    Institute of Scientific and Technical Information of China (English)

    2006-01-01

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

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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. Preliminary modeling of the TMI-2 accident with MELPROG-TRAC

    Energy Technology Data Exchange (ETDEWEB)

    Jenks, R.P.

    1988-01-01

    In support of Nuclear Regulatory Commission and Organization for Economic Cooperation and Development (OECD)-sponsored Three Mile Island-Unit 2 (TMI-2) Analysis Exercise studies, work has been performed to develop a simulation model of the TMI-2 plant for use with the integrated MELPROG-TRAC computer code. Numerous nuclear power plant simulation studies have been performed with the TRAC computer code in the past. Some of these addressed the TMI-2 accident or other hypothetical events at the TMI plant. In addition, studies have been previously performed with the MELPROG-TRAC code using Oconee-1 and Surry plant models. This paper describes the preliminary MELPROG-TRAC input model for severe accident analysis.

  14. 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

  15. Development of LWR Fuels with Enhanced Accident Tolerance

    International Nuclear Information System (INIS)

    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 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% 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 has been carried out for the SiC/SiC composite/SiC monolith structures. A structure with the monolith on the outside and composite on the inside was developed which

  16. The accident consequence model of the German safety study

    International Nuclear Information System (INIS)

    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.)

  17. 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

  18. BRAIN INJURY BIOMECHANICS IN REAL WORLD VEHICLE ACCIDENT USING MATHEMATICAL MODELS

    Institute of Scientific and Technical Information of China (English)

    YANG Jikuang; XU Wei; OTTE Dietmar

    2008-01-01

    This paper aims at investigating brain injury mechanisms and predicting head injuries in real world accidents. For this purpose, a 3D human head finite element model (HBM-head) was developed based on head-brain anatomy. The HBM head model was validated with two experimental tests. Then the head finite element(FE) model and a multi-body system (MBS) model were used to carry out reconstructions of real world vehicle-pedestrian accidents and brain injuries. The MBS models were used for calculating the head impact conditions in vehicle impacts. The HBM-head model was used for calculating the injury related physical parameters, such as intracranial pressure, stress, and strain. The calculated intracranial pressure and strain distribution were correlated with the injury outcomes observed from accidents. It is shown that this model can predict the intracranial biomechanical response and calculate the injury related physical parameters. The head FE model has good biofidelity and will be a valuable tool for the study of injury mechanisms and the tolerance level of the brain.

  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. 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

  1. ATMOSPHERIC MODELING IN SUPPORT OF A ROADWAY ACCIDENT

    Energy Technology Data Exchange (ETDEWEB)

    Buckley, R.; Hunter, C.

    2010-10-21

    The United States Forest Service-Savannah River (USFS) routinely performs prescribed fires at the Savannah River Site (SRS), a Department of Energy (DOE) facility located in southwest South Carolina. This facility covers {approx}800 square kilometers and is mainly wooded except for scattered industrial areas containing facilities used in managing nuclear materials for national defense and waste processing. Prescribed fires of forest undergrowth are necessary to reduce the risk of inadvertent wild fires which have the potential to destroy large areas and threaten nuclear facility operations. This paper discusses meteorological observations and numerical model simulations from a period in early 2002 of an incident involving an early-morning multicar accident caused by poor visibility along a major roadway on the northern border of the SRS. At the time of the accident, it was not clear if the limited visibility was due solely to fog or whether smoke from a prescribed burn conducted the previous day just to the northwest of the crash site had contributed to the visibility. Through use of available meteorological information and detailed modeling, it was determined that the primary reason for the low visibility on this night was fog induced by meteorological conditions.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-06-01

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    OpenAIRE

    Tusheva, Polina

    2013-01-01

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    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)

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

    International Nuclear Information System (INIS)

    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

  12. 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.

  13. An overview of severe accident modeling and analysis work for the ANS reactor conceptual safety analysis report

    International Nuclear Information System (INIS)

    ORNL's Advanced Neutron Source (ANS) will be a new user facility for all kinds of neutron research, centered around a research reactor of unprecedented neutron beam flux. A defense-in-depth philosophy has been adopted. In response to this commitment, ANS Project management has initiated severe accident analysis and related technology development efforts early-on in the design phase itself. Early consideration of severe accident issues will aid in designing a sufficiently robust containment for retention and controlled release of radionuclides in the event of such an accident. It will also provide a means for satisfying on- and off-site regulatory requirements and provide containment response and source term analyses for level-2 and -3 Probabilistic Risk Analyses (PRAs) that will be produced. Moreover, it will provide the best possible understanding of the ANS under severe accident conditions, and consequently provide insights for the development of strategies and design philosophies for accident management, mitigation, and emergency preparedness. This paper presents a perspective overview of the severe accident modeling and analysis work for the ANS Conceptual Safety Analysis Report (CSAR)

  14. 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)

  15. The development and application of the accident dynamic simulator for dynamic probabilistic risk assessment of nuclear power plants

    International Nuclear Information System (INIS)

    This paper describes the principal modelling concepts, practical aspects, and an application of the Accident Dynamic Simulator (ADS) developed for full scale dynamic probabilistic risk assessment (DPRA) of nuclear power plants. Full scale refers not only to the size of the models, but also to the number of potential sequences which should be studied. Plant thermal-hydraulics behaviour, safety systems response, and operator interactions are explicitly accounted for as integrated active parts in the development of accident scenarios. ADS uses discrete dynamic event trees (D-DET) as the main accident scenario modelling approach, and introduces computational techniques to minimize the computer memory requirement and expedite the simulation. An operator model (including procedure-based behaviour and several types of omission and commission errors) and a thermal-hydraulic model with a PC run time more than 300 times faster than real accident time are among the main modules of ADS. To demonstrate the capabilities of ADS, a dynamic PRA of the Steam Generator Tube Rupture event of a US nuclear power plant is analyzed

  16. 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

  17. Co-ordinated research programme on reference studies on probabilistic modelling of accident sequences

    International Nuclear Information System (INIS)

    The co-ordinated research programme (CRP) on probabilistic modelling of accident sequences was established in order to ensure that International Atomic Energy Agency (IAEA) Member States not previously involved in international benchmark exercises obtain adequate practice in applying the available PSA techniques and benefit from the extensive international experience. A supportive peer review group was formed to provide guidance and transfer the insights derived from similar European projects. Seventeen countries participate in this programme which will be completed during 1991. Three working groups have been organized around different reactor types, namely WWER-440 PWRs (with a subgroup analysing AST-500, a district heating plant), Framatome PWRs and CANDU. Each participant in a group studied the same initiating event for a reference plant. For detailed analysis one particular accident sequence has been selected by each team. The logic models (event trees and fault trees) were developed and accident sequences were quantified. Sensitivity analyses are presently in progress. The paper presents some preliminary results and insights. The experiences gained from this CRP are considered as extremely useful for the national PSA programmes in several IAEA Member States. (author). 8 refs, 2 figs, 1 tab

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

    Highlights: • An atmospheric dispersion model for a hypothetical accident of Ghana Research Reactor-1 (GHARR-1) was developed. • Radiological safety analysis after the postulated accident was also carried out. • The MCNPX and HotSpot codes were used to achieve the objectives of our study. • All the values of effective dose obtained following the accident were far below the regulatory limits. - Abstract: Atmospheric dispersion modeling and radiological safety analysis were performed for a postulated accident scenario of the generic Low-Enriched Uranium (LEU) Ghana Research Reactor-1 (GHARR-1) core. The source term was generated from an inventory of peak radioisotope activities released by using the isotope generation code MCNPX. The health physics code, HotSpot, was used to perform the atmospheric transport modeling which was then applied to calculate the total effective dose and how it would be distributed to human organs as a function of distance downwind. All accident scenarios were selected from the GHARR-1 Safety Analysis Report (SAR), assuming that the activities were released to the atmosphere after a design basis accident. The adopted methodology was the use of predominant site-specific meteorological data and dispersion modeling theories to analyze the incident of a hypothetical release to the environment of some selected radionuclides from the site and evaluate to what extent such a release may have radiological effects on the public. The results indicate that all the values of Effective dose obtained, with the maximum of 2.62 × 10−2 mSv at 110 m from the reactor, were far below the regulatory limits, making the use of the reactor safe, even in the event of severe accident scenario

  20. Predictive accident modeling approach inrelation to workover systems

    OpenAIRE

    Jermstad, Lene Bøkseth

    2011-01-01

    Hydro carbon releases are the main contributor to the major accident risk on oil and gas platforms, and the Petroleum Safety Authority Norway (PSA) has thus set a target for reducing such releases. Traditionally topside equipment has been the main focus of study in risk analysis, but to obtain the reduction goals it is important to focus on drilling and well intervention as well. This is due to the complexity of the systems, and the lessons learned from several accidents during such operation...

  1. 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.

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

    International Nuclear Information System (INIS)

    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

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

  4. 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.

  5. Multidisciplinary response for a radiological accident developed in Argentina

    International Nuclear Information System (INIS)

    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 137Cs 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)

  6. Innovative approach to modeling accident response of Gravel Gerties

    Energy Technology Data Exchange (ETDEWEB)

    Kramer, M.; McClure, P.; Sullivan, H.

    1997-08-01

    Recent safety analyses at nuclear explosive facilities have renewed interest in the accident phenomenology associated with explosions in nuclear explosive cells, which are commonly referred to as {open_quotes}Gravel Gerties.{close_quotes} The cells are used for the assembly and disassembly of nuclear explosives and are located in the Device Assembly Facility (DAF) at the Nevada Test Site (NTS) and at the Pantex facility. The cells are designed to mitigate the release of special nuclear material to the environment in the event of a detonation of high explosive within the Gravel Gertie. Although there are some subtle differences between the cells of DAF and Pantex, their general design, geometry, and configuration are similar. The cells consist of a round room approximately 10.4 m in diameter and 5.2 m high enclosed by 0.3-m-thick concrete. Each cell has a wire-rope cantenary roof overlain with gravel. The gravel is approximately 6.9 m deep at the center of the roof and decreases toward the outer edge of the cell. The cell is connected to a corridor and subsequent rooms through an interlocking blast door. In the event of a accidental explosion involving significant amounts of high explosive, the roof structure is lifted by the force of the explosion, the supporting cables break, the gravel is lifted by the blast (resulting in rapid venting of the cell), and the gravel roof collapses, filling the cell. The lifting and subsequent collapse of the gravel, which acts much like a piston, is very challenging to model.

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

    Energy Technology Data Exchange (ETDEWEB)

    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.

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

    International Nuclear Information System (INIS)

    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

  9. Skills development at a paramedic accident simulation centre.

    Science.gov (United States)

    Donaghy, John

    2016-02-01

    Practice simulation in acute and pre-hospital care settings is a growing area of interest for clinicians and health educationalists, and there is much evidence to support its use (Pike and O'Donnell 2010). Most simulation is delivered through computer-aided software or in virtual environments, however last year the University of Hertfordshire opened an accident simulation centre which is an outdoor facility that offers pre- and post-registration paramedics the opportunity to experience a range of scenarios in a 'real life' but secure environment. This article describes how the centre enables students to apply theory to practice in complex situations, such as managing patients injured in road traffic collisions. PMID:26853672

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

    International Nuclear Information System (INIS)

    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

  11. 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.

  12. Speed Spatial Distribution Models for Traffic Accident Section of Freeway Based on Computer Simulation

    Institute of Scientific and Technical Information of China (English)

    Decai Li; Jiangwei Chu; Wenhui Zhang; Xiaojuan Wang; Guosheng Zhang

    2015-01-01

    Simulation models for accident section on freeway are built in microscopic traffic flow simulation environment. In these models involving 2⁃lane, 3⁃lane and 4⁃lane freeway, one detector is set every 10 m to measure section running speed. According to the simulation results, speed spatial distribution curves for traffic accident section on freeway are drawn which help to determine dangerous sections on upstream of accident section. Furthermore, the speed spatial distribution models are obtained for every speed distribution curve. The results provide theoretical basis for determination on temporal and spatial influence ranges of traffic accident and offer reference to formulation of speed limit scheme and other management measures.

  13. Ontology Development and Evolution in the Accident Investigation Domain

    Science.gov (United States)

    Carvalho, Robert; Berrios, Dan; Williams, James

    2004-01-01

    InvestiigationOrganizer (IO) is a collaborative semantic web system designed to support the conduct of mishap investigations. IO provides a common repository for a wide range of mishap related information, allowing investigators to integrate evidence, causal models, and investigation results. IO has been used to support investigations ranging from a small property damage case to the loss of the Space Shuttle Columbia. Through IO'S use in these investigations, we have learned significant lessons? about the application of ontologies and semantic systems to solving real-world problems. This paper will describe the development of the ontology within IO, from the initial development, its growth in response to user requests during use in investigations, and the recent work that was done to control the results of that growth. This paper will also describe the lessons learned from this experience and how they may apply to the implementaton of future ontologies and semantic systems.

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

    International Nuclear Information System (INIS)

    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)

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

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2011-07-01

    .3), FP 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)

  16. 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)

    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)

  17. Radiological accident in Goiania seven years later: what developing country must never forget

    International Nuclear Information System (INIS)

    In September 1987, the removal of the rotating assembly of the shielding head of a teletherapy unit and the dismantling of the capsule containing 50.9 TBq (1375 Ci) of Cs-137 was the most serious radiological accident to have occurred to date. It resulted in the injure by radiation of many people, four of them fatally, and in a widespread contamination of central Goiania, a Brazilian city, of one million inhabitants, Capital of the State of Goias, 180 km far from Brasilia, capital of Brazil. This accident presents lessons that cover the pre-accident period, the emergency phase, and the post-accident phase up to the the present, seven years later, now involving social and economical deliberation for the final Goiania Waste Repository. In each of them were found errors, omissions and correct actions form individuals and organization and, specially in developing country, must be continuously analyzed in terms of safety culture problems identification and safety culture and human beahvior because, not with standing the recommendations contained in publications concerning emergency planning and preparedness, this radiological accident gave evidence of several adverse vectors not mentioned in the literature. Not only social, political, economic and technical problems we had to deal with, but also psychological aspects as fear and depression of the population and also discrimination against the victioms and the main products of the city. This paper also concentrates on safety culture characteristics from the accident lessons, which in developing country are substantially different from the developed one, including the influence and difference of legislation and rules applied, the interfaces and conflicts between the state and the BNCNE and the different aspects involved in providing information and reports to the government, to the organizations and to the community; and finally call the attention from the national and international organizations for questions still in

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

    Energy Technology Data Exchange (ETDEWEB)

    Heames, T.J. (Science Applications International Corp., Albuquerque, NM (USA)); Williams, D.A.; Johns, N.A.; Chown, N.M. (UKAEA Atomic Energy Establishment, Winfrith (UK)); Bixler, N.E.; Grimley, A.J. (Sandia National Labs., Albuquerque, NM (USA)); Wheatley, C.J. (UKAEA Safety and Reliability Directorate, Culcheth (UK))

    1990-10-01

    This document provides a description of a model of the radionuclide behavior in the reactor coolant system (RCS) of a light water reactor during a severe accident. This document serves as the user's manual for the computer code called VICTORIA, based upon the model. The VICTORIA code predicts fission product release from the fuel, chemical reactions between fission products and structural materials, vapor and aerosol behavior, and fission product decay heating. This document provides a detailed description of each part of the implementation of the model into 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. The VICTORIA code was developed upon a CRAY-XMP at Sandia National Laboratories in the USA and a CRAY-2 and various SUN workstations at the Winfrith Technology Centre in England. 60 refs.

  19. A Quasi-Poisson Approach on Modeling Accident Hazard Index for Urban Road Segments

    Directory of Open Access Journals (Sweden)

    Lu Ma

    2014-01-01

    Full Text Available In light of the recently emphasized studies on risk evaluation of crashes, accident counts under specific transportation facilities are adopted to reflect the chance of crash occurrence. The current study introduces more comprehensive measure with the supplement information of accidental harmfulness into the expression of accident risks which are also named Accident Hazard Index (AHI in the following context. Before the statistical analysis, datasets from various sources are integrated under a GIS platform and the corresponding procedures are presented as an illustrated example for similar analysis. Then, a quasi-Poisson regression model is suggested for analyses and the results show that the model is appropriate for dealing with overdispersed count data and several key explanatory variables were found to have significant impact on the estimation of AHI. In addition, the effect of weight on different severity levels of accidents is examined and the selection of the weight is also discussed.

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

    OpenAIRE

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

    2016-01-01

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

  1. Development of accident frequency analysis S/W for chemical processes

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Jae Min; Ko, Jae wook [College of Chemical Engineering, Kwangwoon University (Korea); Shin, Dong Il [School of Chemical Engineering, Seoul National University, Seoul (Korea)

    1999-12-01

    In this study, a computerized prototype program was developed with frequency analysis system as a main system and data base as sub-items to utilize data. Through use of gate-by-gate analysis and minimal cut set using boolean algebra, the frequency analysis program performed the qualitative approach for the accident development path and a quantitative risk analysis. In conclusion, it is thought that the resulting installation will be effective for lessening the probability of accidents through use of this lower cost software. 7 refs., 7 figs.

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

    Energy Technology Data Exchange (ETDEWEB)

    Cantrel, L., E-mail: laurent.cantrel@irsn.fr; Cousin, F.; Bosland, L.; Chevalier-Jabet, K.; Marchetto, C.

    2014-06-01

    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.

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

    International Nuclear Information System (INIS)

    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

  5. 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.

  6. 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.

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

    Directory of Open Access Journals (Sweden)

    Min Yoo

    2014-01-01

    Full Text Available Reliable information through instrumentation systems is essential in mitigating severe accidents such as the one that occurred at the Fukushima Daiichi nuclear power plant. There are five elements which might pose a potential threat to the reliability of parameter detection at nuclear power plants during a severe accident: high temperature, high pressure, high humidity, high radiation, and missiles generated during the evolution of a severe accident. Of these, high temperature apparently poses the most serious threat, since thin shielding can get rid of pressure, humidity, radiation (specifically, alpha and beta radiations, and missile effects. In view of this fact, our study focused on designing an instrument transmitter protecting device that can eliminate the high-temperature effect on transmitters to maintain their functional integrity. We present herein a novel concept for designing such a device in terms of heat transfer model that takes into account various heat transfer mechanisms associated with the device.

  8. Trends in state-level freight accident rates: An enhancement of risk factor development for RADTRAN

    International Nuclear Information System (INIS)

    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

  9. 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. PMID:26915116

  10. Development of modified voxel phantoms for the numerical dosimetric reconstruction of radiological accidents involving external sources: implementation in SESAME tool

    Energy Technology Data Exchange (ETDEWEB)

    Courageot, Estelle; Sayah, Rima; Huet, Christelle [External Dosimetry Department, Institute for Radiological Protection and Nuclear Safety (IRSN), Radiological Protection and Human Health Division, Ionizing Radiation Dosimetry Laboratory, IRSN/DRPH/SDE, BP 17, 92262 Fontenay-aux-Roses Cedex (France)], E-mail: estelle.courageot@irsn.fr

    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. (note)

  11. 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)

    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)

  12. ACCIDENT PREDICTION METHODOLOGY USING CONFLICT ZONE METHOD FOR “TRANSIT TRANSPORT-PEDESTRIAN” CONFLICT SITUATION AND MODELS OF TRAFFIC FLOWS AT CONTROLLED INTERSECTION

    Directory of Open Access Journals (Sweden)

    D. V. Kapsky

    2015-01-01

    Full Text Available Accidents are considered as the most significant cost of road traffic. Therefore any measures for road traffic management should be evaluated according to a minimization  criterion of accident losses. In order to develop a method for evaluation of the accident losses it is necessary to prepare a methodology for cost estimate of road accidents of various severity with due account of their consequences and prediction (economic assessment and severity level of their consequences (quantitative risk assessment. The research has been carried with the purpose to devise appropriate models for accident prediction at a decision-making stage while organizing road traffic in respect of  the “transport-pedestrian” conflict. An interaction of pedestrian and transit road traffic flows  is characterized by rather high risk level. In order to reduce number of road accidents  and  severity of their consequences in the observed conflict, it is necessary to evaluate  proposed solutions, in other words to predict accidents at the stage of object designing and  development of measures.The paper presents its observations on specificity of road traffic and pedestrian flow interactions and analysis of spatial conflict point formation and conflict zone creation in the studied conflict between transport facilities and pedestrians at controlled pedestrian crossings which are located in the area of intersections. Methodology has been developed for accident prediction in accordance with the conflict zone method for various traffic modes at intersections. Dependences of the represented road traffic accidents (according to consequence severity on potential danger of conflicts have been determined for various traffic modes and various conditions of conflict interaction.

  13. 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.)

  14. Development of the source term PIRT based on findings during Fukushima Daiichi NPPs accident

    Energy Technology Data Exchange (ETDEWEB)

    Suehiro, Shoichi, E-mail: suehiro-shouichi@tepsys.co.jp [TEPCO SYSTEMS Co., 2-37-28 Eitai, Koto-Ku, Tokyo 135-0034 (Japan); Sugimoto, Jun [Kyoto University, Yoshida Sakyo, Kyoto 606-8501 (Japan); Hidaka, Akihide [Japan Atomic Energy Agency, Tokai-mura, Naka-gun, Ibaraki 319-1195 (Japan); Okada, Hidetoshi [The Institute of Applied Energy, 14-2 Nishi-Shimbashi 1-Chome, Minato-ku, Tokyo 105-0003 (Japan); Mizokami, Shinya [Tokyo Electric Power Company, 1-3 Uchisaiwai-cho 1-Chome, Chiyoda-ku, Tokyo 100-8560 (Japan); Okamoto, Koji [The University of Tokyo, 2-22 Shirakata, Tokai-mura, Ibaraki 319-1188 (Japan)

    2015-05-15

    Highlights: • We developed the source term PIRT based on findings during the Fukushima accident. • The FoM is the masses or fractions of radionuclides released into the environment. • 68 phenomena were identified as influencing to the FoM. • Radionuclide release from molten fuel had the highest score in the early phase. • MCCI, iodine chemistry, and chemical form had the highest score in the later phase. - Abstract: Research Expert Committee on Evaluation of Severe Accident of AESJ (Atomic Energy Society of Japan) has developed thermal hydraulic PIRT (Phenomena Identification and Ranking Table) and source term (ST) PIRT based on findings during the Fukushima Daiichi NPPs accident. These PIRTs aim 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 analytical codes. The ST PIRT is divided into 3 phases for time domain and 9 categories for spatial domain. The 68 phenomena have been extracted and the importance from the viewpoint of the source term has been ranked through brainstorming and discussions among experts. The present paper describes the developed ST PIRT list and summarizes the high ranked phenomena in each phase.

  15. 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

    2016-01-01

    Background: 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. Objectives: The aim of this study was to develop a minimum data set (MDS) for an information management system to study traffic accidents in Iran. Materials and Methods: 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. Results: 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. Conclusions: 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. PMID:27247791

  16. 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.

  17. Robot developments for the nuclear power accident in TEPCO

    International Nuclear Information System (INIS)

    Many robots had been very active at Fukushima Daiichi Nuclear Power Station since March 11, 2011. Working robots consisted of existing robots such as Packbot firstly entering reactor building and newly developed robots after March 11. Deployment of existing robots required not easy processes such as contract with robot manufacturer, operation training, actual needs and necessary modifications, mockup testing, procedure manual preparation, and real application at site. This article focused on small size robot development in TEPCO, for which the author had been working as remote control technology development team. Robot development started with introduction of DOE-Talon with wired dose ratemeter, and gamma camera and radiation resistant cameras installed on the robot operation vehicle with shielding for high dose measurement around reactor building. Remote technology application committee was set up in June to discuss future needs of robot development and listed required robots corresponding to short-term issues. Development of wired Quince with improved crawler function going up or down stairs had been done with many modifications including cable handling with mockup testing and application experiences. Small size robots for basement survey had been developed for investigation around catwalk in torus room of basement (Survey Runner) and around containment penetration at first floor of unit 1 (FRIGO-MA). Small size robot for high-access survey had been developed for investigation around containment penetration at a high place of first floor of unit 2. (T. Tanaka)

  18. Dynamic modeling of physical phenomena for probabilistic assessment of spent fuel accidents

    Energy Technology Data Exchange (ETDEWEB)

    Benjamin, A.S.

    1997-11-01

    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.

  19. Severe accident modeling of a PWR core with different cladding materials

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, S. C. [Westinghouse Electric Company LLC, 5801 Bluff Road, Columbia, SC 29209 (United States); Henry, R. E.; Paik, C. Y. [Fauske and Associates, Inc., 16W070 83rd Street, Burr Ridge, IL 60527 (United States)

    2012-07-01

    The MAAP v.4 software has been used to model two severe accident scenarios in nuclear power reactors with three different materials as fuel cladding. The TMI-2 severe accident was modeled with Zircaloy-2 and SiC as clad material and a SBO accident in a Zion-like, 4-loop, Westinghouse PWR was modeled with Zircaloy-2, SiC, and 304 stainless steel as clad material. TMI-2 modeling results indicate that lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would result if SiC was substituted for Zircaloy-2 as cladding. SBO modeling results indicate that the calculated time to RCS rupture would increase by approximately 20 minutes if SiC was substituted for Zircaloy-2. Additionally, when an extended SBO accident (RCS creep rupture failure disabled) was modeled, significantly lower peak core temperatures, less H 2 (g) produced, and a smaller mass of molten material would be generated by substituting SiC for Zircaloy-2 or stainless steel cladding. Because the rate of SiC oxidation reaction with elevated temperature H{sub 2}O (g) was set to 0 for this work, these results should be considered preliminary. However, the benefits of SiC as a more accident tolerant clad material have been shown and additional investigation of SiC as an LWR core material are warranted, specifically investigations of the oxidation kinetics of SiC in H{sub 2}O (g) over the range of temperatures and pressures relevant to severe accidents in LWR 's. (authors)

  20. Usefulness of high resolution coastal models for operational oil spill forecast: the Full City accident

    Science.gov (United States)

    Broström, G.; Carrasco, A.; Hole, L. R.; Dick, S.; Janssen, F.; Mattsson, J.; Berger, S.

    2011-06-01

    Oil spill modeling is considered to be an important decision support system (DeSS) 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 is 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 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 including an analysis based on a higher resolution model (1.5 km resolution) for the area the model system show results that compare well with observations. The study also shows that an ensemble using three different models is useful when predicting/analyzing oil spill in coastal areas.

  1. A significant outcome of work life: occupational accidents in a developing country, Turkey.

    Science.gov (United States)

    Ergör, O Alp; Demiral, Yücel; Piyal, Y Bülent

    2003-01-01

    Occupational accident rates are high in the developing world and they are among the crucial indicators of work life. The data on serious occupational accidents in three years from the Aegean Region of Turkey were analyzed in order to define the working life situation. The data for 1995-97 were obtained from The Social Insurance Institution. Variables were defined for workers, work environment and working conditions for descriptive analysis. The relationships between these variables and outcomes of the occupational accidents were analyzed. The fatality risk was 2.6 times higher for mining and construction than for the production sector (p<0.001), and 1.6 times higher for evening and night shifts than for day shifts (p<0.01). Mortality from being struck by something, falls, motor vehicle accidents and electrocutions were significantly higher than being cut or in, under or between objects (OR 3.5, 8.7, 23.1, 26.1 respectively and p<0.01). As a conclusion, it was suggested that within the globalization picture, occupational health problems in the developing world should be explored intensively in order to establish effective prevention programs. PMID:14605434

  2. 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

  3. 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

  4. Usefulness of high resolution coastal models for operational oil spill forecast: the "Full City" accident

    Science.gov (United States)

    Broström, G.; Carrasco, A.; Hole, L. R.; Dick, S.; Janssen, F.; Mattsson, J.; Berger, S.

    2011-11-01

    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.

  5. 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.

  6. 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)

    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. Al3Ti-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 ceramic

  7. Health effects models for nuclear power plant accident consequence analysis: Low LET radiation

    International Nuclear Information System (INIS)

    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

  8. 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.

  9. Development of NASA's Accident Precursor Analysis Process Through Application on the Space Shuttle Orbiter

    Science.gov (United States)

    Maggio, Gaspare; Groen, Frank; Hamlin, Teri; Youngblood, Robert

    2010-01-01

    Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system. APA docs more than simply track experience: it systematically evaluates experience, looking for under-appreciated risks that may warrant changes to design or operational practice. This paper presents the pilot application of the NASA APA process to Space Shuttle Orbiter systems. In this effort, the working sessions conducted at Johnson Space Center (JSC) piloted the APA process developed by Information Systems Laboratories (ISL) over the last two years under the auspices of NASA's Office of Safety & Mission Assurance, with the assistance of the Safety & Mission Assurance (S&MA) Shuttle & Exploration Analysis Branch. This process is built around facilitated working sessions involving diverse system experts. One important aspect of this particular APA process is its focus on understanding the physical mechanism responsible for an operational anomaly, followed by evaluation of the risk significance of the observed anomaly as well as consideration of generalizations of the underlying mechanism to other contexts. Model completeness will probably always be an issue, but this process tries to leverage operating experience to the extent possible in order to address completeness issues before a catastrophe occurs.

  10. An approach to modelling operator behaviour in integrated dynamic accident sequence analysis

    International Nuclear Information System (INIS)

    The paper describes an integrated dynamic methodology for simulating nuclear power plant accidents, with special focus on the operator behaviour model. The overall model consists of accident sequence pre-processor, operator response model, safety and support system model, plant dependence model, thermal hydraulics model, and accident sequence scheduler. The operator model consists of the knowledge base (KB) and the decision making module (DM). KB consists of rules of behaviour. Behaviour is guided by emergency operating procedures (EOPs), thermal hydraulics parameters of the plant, system status, and other factors including stress, training, experience, etc. Possible error mechanisms in following symptom based EOPs are mentioned, and factors which cause some of these errors are identified. Plant parameters are classified as ''diagnostic'' and ''control''. Comparison of operator expectations and plant inputs guides the behaviour. System states affect only control action and not diagnosis. The decision maker simulates the operator behaviour in the way it accesses the KB, assuming that the KB contains all the knowledge that is necessary for managing the accident. This is modelled through a ''filter'' concept where the factors that affect behaviour are filters that affect the access to KB. Actions are categorized in verifying the response of reactor protection systems, and in controlling inventory and heat removal. System modelling is done at system rather than component level since operator actions affect the plant at system level. The methodology is being implemented in PC environment. Possible applications include analysis of causes and consequences of operator actions, particularly errors of commission, EOP validation, analysis of dynamic effects of accident sequences, and performing probabilistic risk assessments. 15 refs, 2 figs, 1 tab

  11. Fuel Behaviour and Modelling under Severe Transient and Loss of Coolant Accident (LOCA) Conditions. Proceedings of a Technical Meeting

    International Nuclear Information System (INIS)

    In recent years the demands on 'fuel duties' have increased, including transient regimes, higher burnups and longer fuel cycles. To satisfy these demands, fuel vendors have developed and introduced new cladding and fuel material designs to provide sufficient margins for safe operation of the fuel components. National and international experimental programmes have been launched, and models have been developed or adapted to take into account the changed conditions. These developments enable water cooled reactors, which contribute about 95% of the nuclear power in the world today, to operate safely under all operating conditions; moreover, even under severe transient or accident conditions, such as reactivity initiated accidents (RIAs) or loss of coolant accidents (LOCAs), the behaviour of the fuel can be adequately predicted and the consequences of such events can be safely contained. In 2010 the IAEA Technical Working Group on Fuel Performance and Technology (TWGFPT) recommended that a technical meeting on ''Fuel Behaviour and Modelling under Severe Transient and LOCA Conditions'' be held in Japan. The accident at the Fukushima Daiichi nuclear power plant in March 2011 highlighted the need to address this subject, and despite the difficult situation in Japan at the time, the recommended plan was confirmed, and the Japan Atomic Energy Agency (JAEA) hosted the technical meeting in Mito, Ibaraki Prefecture, Japan, from 18 to 21 October 2011. This meeting was the eighth in a series of IAEA meetings, which reflects Member States' continuing interest in the above issues. The previous meetings were held in 1980 (jointly with OECD Nuclear Energy Agency, Helsinki, Finland), 1983 (Riso, Denmark), 1986 (Vienna, Austria), 1988 (Preston, United Kingdom), 1992 (Pembroke, Canada), 1995 (Dimitrovgrad, Russian Federation) and 2001 (Halden, Norway). The purpose of the technical meeting was to provide a forum for international experts to review the current situation and the state of

  12. WASTE-ACC: A computer model for analysis of waste management accidents

    Energy Technology Data Exchange (ETDEWEB)

    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.

  13. 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

  14. Development of the post-accident expert assessment capacity of I.R.S.N

    International Nuclear Information System (INIS)

    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)

  15. A dynamic event tree informed approach to probabilistic accident sequence modeling: Dynamics and variabilities in medium LOCA

    International Nuclear Information System (INIS)

    In Probability Safety Assessments, accident scenario dynamics are addressed in the accident sequence analysis task. In an analyst-driven, iterative process, assumptions are made about equipment responses and operator actions and simulations of the scenario evolution are performed. To calculate how scenario dynamics and stochastic variabilities may affect the results of this process in terms of estimated risk, this work applies Dynamic Event Trees (DETs) to more comprehensively examine the accident scenario space. Alternative event tree models are developed and the core damage frequency is quantified to reveal the effects of different delineations of the sequences and of the bounding assumptions underlying success criteria. The results from a case study on Medium-break Loss of Coolant Accident scenarios in a Pressurized Water Reactor are presented, considering the break size, available injection trains, and the timing of rapid cooldown and the switchover to recirculation. The results show not only that estimated risk can be very sensitive to the numerous assumptions made in current accident sequence analysis but also that bounding assumptions do not always result in conservative risk estimates, thereby confirming the benefits that DETs provide in terms of characterizing scenario dynamics. - Highlights: • The overall most challenging MLOCA break is at neither extreme of the size range. • Selecting the limiting break size influenced estimated risk strongly (6″ vs 7″). • Success criteria can be defined more realistically by splitting the MLOCA range. • A more demanding success criterion for one top event can reduce overall risk. • Non-limiting success branches may lead to more demanding subsequent success criteria

  16. Cold leg condensation model for analyzing loss-of-coolant accident in PWR

    Energy Technology Data Exchange (ETDEWEB)

    Liao, Jun, E-mail: liaoj@westinghouse.com; Frepoli, Cesare; Ohkawa, Katsuhiro

    2015-04-15

    Highlights: • Direct contact cold leg condensation model for full spectrum LOCA evaluation model. • The cold leg condensation model addresses both large break LOCA and small break LOCA. • The model is assessed against both large break and small break LOCA experiments. • Scalability of the cold leg condensation model to full scale PWR is discussed. - Abstract: Direct contact condensation in the cold leg of pressurized water reactor is an important phenomenon during a postulated loss-of-coolant accident. The amount of condensation in the cold legs impacts the thermal hydraulic behavior of the reactor coolant system and eventually the integration of reactor nuclear core. A cold leg condensation model was developed for the WCOBRA/TRAC-TF2 safety analysis code. The model correlated the COSI test data and addressed the scaling issues with respect to geometry, pressure, and steam and water flow rates expected during a typical PWR LOCA. The correlation was found to be in good agreement with separate effects and integral effects experimental data and implemented in the WCOBRA/TRAC-TF2 safety analysis code. The cold leg condensation model was assessed against various small break and large break LOCA separate effects tests such as COSI experiments, ROSA experiments and UPTF experiments. Those experiments cover a wide range of cold leg dimensions, system pressures, mass flow rates, and fluid properties. All the predicted condensation results match reasonably well with the experimental data. Scalability discussions on the diameter, flow area, length, superficial velocity, Reynolds number of both cold leg and SI line, and Froude number of SI line in the Westinghouse COSI test facility were provided. The distortion of the SI jet Reynolds number is moderate. The scaling analysis together with the validation matrix covering a wide range of cold leg diameter, SI flow rate and SI Reynolds number support the scalability of the developed cold leg condensation model to the full

  17. 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

    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...... 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...

  18. A two-stage optimization model for emergency material reserve layout planning under uncertainty in response to environmental accidents.

    Science.gov (United States)

    Liu, Jie; Guo, Liang; Jiang, Jiping; Jiang, Dexun; Liu, Rentao; Wang, Peng

    2016-06-01

    In the emergency management relevant to pollution accidents, efficiency emergency rescues can be deeply influenced by a reasonable assignment of the available emergency materials to the related risk sources. In this study, a two-stage optimization framework is developed for emergency material reserve layout planning under uncertainty to identify material warehouse locations and emergency material reserve schemes in pre-accident phase coping with potential environmental accidents. This framework is based on an integration of Hierarchical clustering analysis - improved center of gravity (HCA-ICG) model and material warehouse location - emergency material allocation (MWL-EMA) model. First, decision alternatives are generated using HCA-ICG to identify newly-built emergency material warehouses for risk sources which cannot be satisfied by existing ones with a time-effective manner. Second, emergency material reserve planning is obtained using MWL-EMA to make emergency materials be prepared in advance with a cost-effective manner. The optimization framework is then applied to emergency management system planning in Jiangsu province, China. The results demonstrate that the developed framework not only could facilitate material warehouse selection but also effectively provide emergency material for emergency operations in a quick response.

  19. 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.

  20. A review of accident response models for risk assessments involving the transport of spent nuclear fuel

    International Nuclear Information System (INIS)

    A study was performed to explore the differences between two spent fuel transportation risk assessment models used to calculate conditional accident probabilities and radionuclide release fractions. The Wilmot model, from work performed at Sandia National Laboratories, and the NRC-sponsored Modal Study model were compared to identify areas of conservatism and to assess their applicability to current risk assessment studies. The study included reviewing model assumptions, mathematical equations, and data sources for each model. The total probability hazard results showed that Modal Study gave several orders of magnitude higher total relative risk than the Wilmot values. However, considering the very low magnitudes of the risk, this difference is not considered significant with respect to the overall risk assessment. It was also found that the documentation and referencing of accident response region models needs improvements

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

    International Nuclear Information System (INIS)

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

  2. System analysis with improved thermo-mechanical fuel rod models for modeling current and advanced LWR materials in accident scenarios

    Science.gov (United States)

    Porter, Ian Edward

    A nuclear reactor systems code has the ability to model the system response in an accident scenario based on known initial conditions at the onset of the transient. However, there has been a tendency for these codes to lack the detailed thermo-mechanical fuel rod response models needed for accurate prediction of fuel rod failure. This proposed work will couple today's most widely used steady-state (FRAPCON) and transient (FRAPTRAN) fuel rod models with a systems code TRACE for best-estimate modeling of system response in accident scenarios such as a loss of coolant accident (LOCA). In doing so, code modifications will be made to model gamma heating in LWRs during steady-state and accident conditions and to improve fuel rod thermal/mechanical analysis by allowing axial nodalization of burnup-dependent phenomena such as swelling, cladding creep and oxidation. With the ability to model both burnup-dependent parameters and transient fuel rod response, a fuel dispersal study will be conducted using a hypothetical accident scenario under both PWR and BWR conditions to determine the amount of fuel dispersed under varying conditions. Due to the fuel fragmentation size and internal rod pressure both being dependent on burnup, this analysis will be conducted at beginning, middle and end of cycle to examine the effects that cycle time can play on fuel rod failure and dispersal. Current fuel rod and system codes used by the Nuclear Regulatory Commission (NRC) are compilations of legacy codes with only commonly used light water reactor materials, Uranium Dioxide (UO2), Mixed Oxide (U/PuO 2) and zirconium alloys. However, the events at Fukushima Daiichi and Three Mile Island accident have shown the need for exploration into advanced materials possessing improved accident tolerance. This work looks to further modify the NRC codes to include silicon carbide (SiC), an advanced cladding material proposed by current DOE funded research on accident tolerant fuels (ATF). Several

  3. The usefulness of time-dependent reactor accident consequence modelling for emergency response planning

    International Nuclear Information System (INIS)

    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

  4. Generation IV benchmarking of TRISO fuel performance models under accident conditions: Modeling input data

    Energy Technology Data Exchange (ETDEWEB)

    Collin, Blaise P. [Idaho National Laboratory (INL), Idaho Falls, ID (United States)

    2014-09-01

    This document presents the benchmark plan for the calculation of particle fuel performance on safety testing experiments that are representative of operational accidental transients. The benchmark is dedicated to the modeling of fission product release under accident conditions by fuel performance codes from around the world, and the subsequent comparison to post-irradiation experiment (PIE) data from the modeled heating tests. The accident condition benchmark is divided into three parts: the modeling of a simplified benchmark problem to assess potential numerical calculation issues at low fission product release; the modeling of the AGR-1 and HFR-EU1bis safety testing experiments; and, the comparison of the AGR-1 and HFR-EU1bis modeling results with PIE data. The simplified benchmark case, thereafter named NCC (Numerical Calculation Case), is derived from ''Case 5'' of the International Atomic Energy Agency (IAEA) Coordinated Research Program (CRP) on coated particle fuel technology [IAEA 2012]. It is included so participants can evaluate their codes at low fission product release. ''Case 5'' of the IAEA CRP-6 showed large code-to-code discrepancies in the release of fission products, which were attributed to ''effects of the numerical calculation method rather than the physical model''[IAEA 2012]. The NCC is therefore intended to check if these numerical effects subsist. The first two steps imply the involvement of the benchmark participants with a modeling effort following the guidelines and recommendations provided by this document. The third step involves the collection of the modeling results by Idaho National Laboratory (INL) and the comparison of these results with the available PIE data. The objective of this document is to provide all necessary input data to model the benchmark cases, and to give some methodology guidelines and recommendations in order to make all results suitable for comparison

  5. 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

  6. 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.

  7. 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.

  8. 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)

    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

  9. Calculation notes that support accident scenario and consequence development for the leak from a railcar/tank trailer at the 204-ar waste unloading facility

    Energy Technology Data Exchange (ETDEWEB)

    Ryan, G.W., Westinghouse Hanford

    1996-09-19

    This document supports the development and presentation of the following accident scenario in the TWRS Final Safety Analysis Report: Leak from Railcar/Tank Trailer. The calculations needed to quantify the risk associated with this accident scenario are included within.

  10. Characteristics of the development of the radiological situation resulting from the accident, intervention levels and countermeasures

    International Nuclear Information System (INIS)

    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

  11. Development of a MAAP-based Severe Accident Training Simulator using Visual System Analyzer

    Energy Technology Data Exchange (ETDEWEB)

    Suh, Jae Seung [SENTECH, Daejeon (Korea, Republic of); Park, Soo Yong; Ahn, Kwang Il; Kim, Kyung Doo [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2011-10-15

    The recent environment of severe accident analysis requires high performance computers to simulate complicated reactor and containment phenomena. In parallel with this, rapid advances in computer technology now enable these codes to run in real or almost real time. The remaining limitation restricting their use on an even wider scale is that most of the existing codes are still subject to a complicated I/O structure. Even user-friendly graphical user interfaces (GUI) will be not likely to help better and efficient interpretation of the analysis results obtained from these codes as well as for their increased use. This situation has motivated the development of easy-to-use GUI tools for severe accident codes, such as ViSA, SNAP, MAAP4-GRAAP, SATS, and et al. For instance, ViSA enables the thermal-hydraulic system codes to be used like a conventional nuclear plant analyzer. Recently, a project for a real time simulation of results obtained using MAAP4 codes under the ViSA environment has been initiated in KAERI. Such a GUI-based interactive interface can be very useful in sharing real time analysis results obtained from the MAAP code. The purpose of this paper is to introduce the current status of a MAAP-based Severe Accident Simulator being coupled with the ViSA system

  12. Fault-tree Models of Accident Scenarios of RoPax Vessels

    Institute of Scientific and Technical Information of China (English)

    Pedro Ant(a)o; C. Guedes Soares

    2006-01-01

    Ro-Ro vessels for cargo and passengers (RoPax) are a relatively new concept that has proven to be popular in the Mediterranean region and is becoming more widespread in Northern Europe. Due to its design characteristics and amount of passengers, although less than a regular passenger liner, accidents with RoPax vessels have far reaching consequences both for economical and for human life. The objective of this paper is to identify hazards related to casualties of RoPax vessels. The terminal casualty events chosen are related to accident and incident statistics for this type of vessel. This paper focuses on the identification of the basic events that can lead to an accident and the performance requirements. The hazard identification is carried out as the first step of a Formal Safety Assessment (FSA) and the modelling of the relation between the relevant events is made using Fault Tree Analysis (FTA). The conclusions of this study are recommendations to the later steps of FSA rather than for decision making (Step 5 of FSA). These recommendations will be focused on the possible design shortcomings identified during the analysis by fault trees throughout cut sets. Also the role that human factors have is analysed through a sensitivity analysis where it is shown that their influence is higher for groundings and collisions where an increase of the initial probability leads to the change of almost 90% of the accident occurrence.

  13. Uncertainties propagation in the framework of a Rod Ejection Accident modeling based on a multi-physics approach

    International Nuclear Information System (INIS)

    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)

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

    Institute of Scientific and Technical Information of China (English)

    1998-01-01

    A dynamic food-chain model and program, DYFOM-95, forpredicting the radiological consequences of nuclear accident hasbeen developed, which is not only suitable to the West food-chainbut also to Chinese food chain. The following processes, caused byaccident release which will make an impact on radionuclideconcentration in the edible parts of vegetable are considered: dryand wet deposition interception and initial retention,translocation, percolation, root uptake and tillage. Activityintake rate of animals, effects of processing and activity intakeof human through ingestion pathway are also considered incalculations. The effects of leaf area index LAI of vegetable areconsidered in dry deposition model. A method for calculating thecontribution of rain with different period and different intensityto total wet deposition is established. The program contains 1 maincode and 5 sub-codes to calculate dry and wet deposition on surfaceof vegetable and soil, translocation of nuclides in vegetable,nuclide concentration in the edible parts of vegetable and inanimal products and activity intake of human and so on.

  15. Modeling of the TMI-2 (Three Mile Island Unit-2) accident with MELPROG/TRAC and calculation results for Phases 1 and 2

    Energy Technology Data Exchange (ETDEWEB)

    Motley, F.E.; Jenks, R.P.

    1988-01-01

    Work has been performed to develop a Three Mile Island Unit-2 (TMI-2) simulation model for MELPROG/TRAC capable of predicting the observed plant behavior that took place during the accident of March 1979. A description of the TMI-2 plant model is presented and calculation results through 174 min of the accident are discussed. Using the ICBC boundary conditions, the calculation predicts pressurizer draining and core recovering prior to fuel-rod damage. A parametric calculation (reduced makeup flow) is currently underway and is in better agreement with the observed plant behavior. Efforts are underway to resolve current discrepancies and proceed with an accurate simulation through Phases 3 and 4 of the accident (174-227 min and 227-300 min, respectively). 13 refs., 11 figs., 2 tabs.

  16. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    Energy Technology Data Exchange (ETDEWEB)

    Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Johnson, J.D.; Rollstin, J.A. [GRAM, Inc., Albuquerque, NM (United States); Shiver, A.W.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States)

    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.

  17. Uncertainty and sensitivity analysis of food pathway results with the MACCS Reactor Accident Consequence Model

    International Nuclear Information System (INIS)

    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

  18. 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.

  19. Development of heat insulation device to protect pressure measuring instruments from high temperature under the severe accident

    Energy Technology Data Exchange (ETDEWEB)

    Ham, Jaehyun; Shin, Sung Min; Kang, Hyun Gook [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2015-10-15

    Micro Control Unit (MCU), communication module, and power supply system are also needed to be protected for the pressure transmitter. The harsh condition in containment which is created by the severe accident are composed of five elements: high temperature, high pressure, high humidity, high radiation, and physical threats by shrapnel generated during the process of the severe accident. Among these five elements, high temperature should be focused because other elements can be solved even with the thin shield. In this study, a detailed design of the heat insulation device which will be installed in the containment based on the Min Yoo's study and a verification test are done. Development of heat insulation device which enables operator to get in-containment data for the proper mitigation process under the severe accident was done in this study. With researches for severe accident management systems which proceeding actively since the Fukushima accident, researches for reliable instrumentations of in-containment data which is necessary to operate severe accident management systems properly in harsh condition during accident also should be progressed.

  20. Accidents - Chernobyl accident

    International Nuclear Information System (INIS)

    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.)

  1. Development of Co-Pilgering Process for Manufacturing Double Clad Tubes for Accident Tolerant Fuel

    International Nuclear Information System (INIS)

    Accident Tolerant Fuels (ATF) are those that, in comparison with the standard UO2 - Zr system, can tolerate loss of active cooling in the core for a considerably longer time period (depending on the accident scenario), while maintaining or improving the fuel performance during normal operations. ATF cladding development efforts focus on materials with more benign steam reaction. For this, advanced steels (e.g. FeCrAl), refractory metals (e.g. Mo), ceramic cladding (SiC), Innovative alloys with dopants, zirconium alloy with coating or sleeve are being developed. Single material like zirconium alloy as clad may not be compatible with both fuel and coolant at elevated temperatures in accident scenario. Double clad tube is one of the prime concepts which has to be explored to develop ATF cladding. Two different clad materials- one oxidant resistant (like FeCrAl) and the other, fuel compatible (like Zr-4) constitute together as outer and inner tube to form ATF cladding. Bonding two different tubes in controlled thickness ratios and with almost no gap in between is utmost difficult. Different types of processes are available for production of double clad tubes such as coating, co-extrusion, co- drawing, internal expansion/external compaction, explosive bonding, co-pilgering etc,. Nuclear Fuel Complex (NFC), India has successfully demonstrated manufacturing of double clad tube by co-pilgering process where in outer cladding is of modified 9Cr-1Mo Steel and inner liner is of zircaloy-4. Considering different deformation behaviour of above materials during pilgering, fabrication of double clad tube is very critical. Optimization of tube dimensions like outer diameter and wall thickness at pre and final stages during pilgering is very important to achieve the required overall tube dimension and bonding between the tubes. This paper gives the methodology of manufacture of Double Clad Tubes by pilgering and the bonding between the two materials achieved in this process

  2. Anthropometric dependence of the response of a thorax FE model under high speed loading: validation and real world accident replication.

    Science.gov (United States)

    Roth, Sébastien; Torres, Fabien; Feuerstein, Philippe; Thoral-Pierre, Karine

    2013-05-01

    Finite element analysis is frequently used in several fields such as automotive simulations or biomechanics. It helps researchers and engineers to understand the mechanical behaviour of complex structures. The development of computer science brought the possibility to develop realistic computational models which can behave like physical ones, avoiding the difficulties and costs of experimental tests. In the framework of biomechanics, lots of FE models have been developed in the last few decades, enabling the investigation of the behaviour of the human body submitted to heavy damage such as in road traffic accidents or in ballistic impact. In both cases, the thorax/abdomen/pelvis system is frequently injured. The understanding of the behaviour of this complex system is of extreme importance. In order to explore the dynamic response of this system to impact loading, a finite element model of the human thorax/abdomen/pelvis system has, therefore, been developed including the main organs: heart, lungs, kidneys, liver, spleen, the skeleton (with vertebrae, intervertebral discs, ribs), stomach, intestines, muscles, and skin. The FE model is based on a 3D reconstruction, which has been made from medical records of anonymous patients, who have had medical scans with no relation to the present study. Several scans have been analyzed, and specific attention has been paid to the anthropometry of the reconstructed model, which can be considered as a 50th percentile male model. The biometric parameters and laws have been implemented in the dynamic FE code (Radioss, Altair Hyperworks 11©) used for dynamic simulations. Then the 50th percentile model was validated against experimental data available in the literature, in terms of deflection, force, whose curve must be in experimental corridors. However, for other anthropometries (small male or large male models) question about the validation and results of numerical accident replications can be raised.

  3. Economic development and traffic accident mortality in the industrialized world, 1962-1990

    NARCIS (Netherlands)

    E.F. van Beeck (Ed); G.J.J.M. Borsboom (Gerard); J.P. Mackenbach (Johan)

    2000-01-01

    textabstractBACKGROUND: We examined the association between prosperity and traffic accident mortality in the industrialized world in a long-term perspective. METHODS: We calculated traffic accident mortality, traffic mobility and the fatal injury rate of 21 industrializ

  4. 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)

    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

  5. 基于事故发展与控制的隐患分级方法%Risk Classification Method for Accident Potential Based on Development and Control Measures of Accident

    Institute of Scientific and Technical Information of China (English)

    赵东风; 申玉琪; 赵志强; 张佑明; 孟亦飞

    2012-01-01

    For the purpose of making the management of accident potential more scientific, the relationship between accident potential and accident was discussed and studied. The fundamental property of accident potential was brought forward that it can make the accident happen or develop. The mechanism whereby accident potential functions in the accident was revealed clearly, by predicting the impelling and inhibiting factors in the accident process. The accident potential was classified into two types (the first type of accident potential and the second type of accident potential) according to the different time they work. In the risk assessment, the problem of specific accident potential grading was resolved. By introducing the assessment indexes of accident potential exposure frequency, possibility of other factors, corrective actions, the initial value of the consequences of the accident, the correction factor of personnel protection, the correction factor of personnel exposure, the correction factor of emergency measures, the correction factor of property loss, an assessment index system was established. By calculating the risk of accident potential-causing accident, the real risk of accident potential was assessed.%为使隐患管理工作更加科学,对隐患与事故的关系进行研讨,提出隐患的根本属性是能够促使事故发生或发展.通过预估促使和控制(阻碍)事故发展的因素,来揭示隐患在事故过程中的作用机制.根据发生作用的时间将隐患分为第1类隐患和第2类隐患.在风险评估过程中,解决了具体隐患风险分级的问题,提出隐患暴露频率、其他条件的可能性、隐患纠正系数、事故后果初始分值、人员防护修正系数、人员暴露修正系数、应急处理与事故控制修正系数和财产损失修正系数等评价指标.通过隐患致因事故风险的计算,评估隐患的最终风险.

  6. 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)

  7. 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.

  8. 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.

  9. Uncertainty and sensitivity analysis of early exposure results with the MACCS Reactor Accident Consequence Model

    Energy Technology Data Exchange (ETDEWEB)

    Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Johnson, J.D. [GRAM, Inc., Albuquerque, NM (United States); McKay, M.D. [Los Alamos National Lab., NM (United States); Shiver, A.W.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States)

    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.

  10. Models and numerical methods for the simulation of loss-of-coolant accidents in nuclear reactors

    Science.gov (United States)

    Seguin, Nicolas

    2014-05-01

    In view of the simulation of the water flows in pressurized water reactors (PWR), many models are available in the literature and their complexity deeply depends on the required accuracy, see for instance [1]. The loss-of-coolant accident (LOCA) may appear when a pipe is broken through. The coolant is composed by light water in its liquid form at very high temperature and pressure (around 300 °C and 155 bar), it then flashes and becomes instantaneously vapor in case of LOCA. A front of liquid/vapor phase transition appears in the pipes and may propagate towards the critical parts of the PWR. It is crucial to propose accurate models for the whole phenomenon, but also sufficiently robust to obtain relevant numerical results. Due to the application we have in mind, a complete description of the two-phase flow (with all the bubbles, droplets, interfaces…) is out of reach and irrelevant. We investigate averaged models, based on the use of void fractions for each phase, which represent the probability of presence of a phase at a given position and at a given time. The most accurate averaged model, based on the so-called Baer-Nunziato model, describes separately each phase by its own density, velocity and pressure. The two phases are coupled by non-conservative terms due to gradients of the void fractions and by source terms for mechanical relaxation, drag force and mass transfer. With appropriate closure laws, it has been proved [2] that this model complies with all the expected physical requirements: positivity of densities and temperatures, maximum principle for the void fraction, conservation of the mixture quantities, decrease of the global entropy… On the basis of this model, it is possible to derive simpler models, which can be used where the flow is still, see [3]. From the numerical point of view, we develop new Finite Volume schemes in [4], which also satisfy the requirements mentioned above. Since they are based on a partial linearization of the physical

  11. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    International Nuclear Information System (INIS)

    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

  12. Uncertainty and sensitivity analysis of chronic exposure results with the MACCS reactor accident consequence model

    Energy Technology Data Exchange (ETDEWEB)

    Helton, J.C. [Arizona State Univ., Tempe, AZ (United States); Johnson, J.D.; Rollstin, J.A. [Gram, Inc., Albuquerque, NM (United States); Shiver, A.W.; Sprung, J.L. [Sandia National Labs., Albuquerque, NM (United States)

    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.

  13. Application of Structural Equations Modeling to assess relationship among Emotional Intelligence, General Health and Occupational Accidents

    Directory of Open Access Journals (Sweden)

    MOAMMAD KHANDAN

    2015-09-01

    Full Text Available ORIGINAL ARTICLEEmotional intelligence (EI has been subject of significant amounts of literature over the past two decades. However, little has been contributed to how emotional intelligence may be practically applied to enhance both accident prevention program and general health in workplaces. Purpose of this paper is to survey relationship among these variables in working society of Iran in 2014. As well as identify practical approaches to application of emotional intelligence skills to manage work change process.This was a cross-sectional study, conducted among all workers in functional units of a manufacturing company (n=178, located in a central province in Iran. Emotional intelligence assessed using Bradberry and Greaves’ questionnaire and Goldberg’s General Health Questionnaire (GHQ was the other tool used in the study. Descriptive statistics used to describe data by SPSS V22. Also, relationship among studied factors analyzed applying structural equations (SEM modeling by EQS software. Majority of workers (99.32% were male. Mean (SD age was 39.13 (8.23 also 64.19% of participants were married. Mean and standard deviation of EI score calculated 90.64 and 19.33, respectively. Also, results indicated that mean of GH score was 22.24 (±9.83. Analyzing relation between main variables (EI & GH with occupational accidents depicted that both of them are in significant relationship with accident (P<0.05. Regard to relationship between emotional intelligence and general health with occurrence of accidents in workplace and GH improvement with increase in EI, three strategies are recommended: appropriate job selection, Suitable training and intervention for workplace condition improvement.

  14. Light-Weight Radioisotope Heater Unit final safety analysis report (LWRHU-FSAR): Volume 2: Accident Model Document (AMD)

    Energy Technology Data Exchange (ETDEWEB)

    Johnson, E.W.

    1988-10-01

    The purpose of this volume of the LWRHU SAR, the Accident Model Document (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; Provide estimates of occurrence probabilities associated with these various accidents; Evaluate the response of the LWRHU (or its components) to the resultant accident environments; and Associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere.

  15. The development of a nuclear accident risk information system(NARIS)

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, Jong Tae; Jung, Won Dea

    2001-03-01

    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 analysis 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.

  16. The development of a nuclear accident risk information system(NARIS)

    International Nuclear Information System (INIS)

    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 analysis 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

  17. Comparison of the foodchain transport models of WASH-1400 and MARC using the accident consequence model UFOMOD

    International Nuclear Information System (INIS)

    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.)

  18. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    Science.gov (United States)

    Zhang, Yingyu; Shao, Wei; Zhang, Mengjia; Li, Hejun; Yin, Shijiu; Xu, Yingjun

    2016-07-01

    Mining has been historically considered as a naturally high-risk industry worldwide. Deaths caused by coal mine accidents are more than the sum of all other accidents in China. Statistics of 320 coal mine accidents in Shandong province show that all accidents contain indicators of "unsafe conditions of the rules and regulations" with a frequency of 1590, accounting for 74.3% of the total frequency of 2140. "Unsafe behaviors of the operator" is another important contributory factor, which mainly includes "operator error" and "venturing into dangerous places." A systems analysis approach was applied by using structural equation modeling (SEM) to examine the interactions between the contributory factors of coal mine accidents. The analysis of results leads to three conclusions. (i) "Unsafe conditions of the rules and regulations," affect the "unsafe behaviors of the operator," "unsafe conditions of the equipment," and "unsafe conditions of the environment." (ii) The three influencing factors of coal mine accidents (with the frequency of effect relation in descending order) are "lack of safety education and training," "rules and regulations of safety production responsibility," and "rules and regulations of supervision and inspection." (iii) The three influenced factors (with the frequency in descending order) of coal mine accidents are "venturing into dangerous places," "poor workplace environment," and "operator error." PMID:27085591

  19. Analysis 320 coal mine accidents using structural equation modeling with unsafe conditions of the rules and regulations as exogenous variables.

    Science.gov (United States)

    Zhang, Yingyu; Shao, Wei; Zhang, Mengjia; Li, Hejun; Yin, Shijiu; Xu, Yingjun

    2016-07-01

    Mining has been historically considered as a naturally high-risk industry worldwide. Deaths caused by coal mine accidents are more than the sum of all other accidents in China. Statistics of 320 coal mine accidents in Shandong province show that all accidents contain indicators of "unsafe conditions of the rules and regulations" with a frequency of 1590, accounting for 74.3% of the total frequency of 2140. "Unsafe behaviors of the operator" is another important contributory factor, which mainly includes "operator error" and "venturing into dangerous places." A systems analysis approach was applied by using structural equation modeling (SEM) to examine the interactions between the contributory factors of coal mine accidents. The analysis of results leads to three conclusions. (i) "Unsafe conditions of the rules and regulations," affect the "unsafe behaviors of the operator," "unsafe conditions of the equipment," and "unsafe conditions of the environment." (ii) The three influencing factors of coal mine accidents (with the frequency of effect relation in descending order) are "lack of safety education and training," "rules and regulations of safety production responsibility," and "rules and regulations of supervision and inspection." (iii) The three influenced factors (with the frequency in descending order) of coal mine accidents are "venturing into dangerous places," "poor workplace environment," and "operator error."

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

    International Nuclear Information System (INIS)

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

  1. Accident knowledge and emergency management

    Energy Technology Data Exchange (ETDEWEB)

    Rasmussen, B.; Groenberg, C.D.

    1997-03-01

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs.

  2. Accident knowledge and emergency management

    International Nuclear Information System (INIS)

    The report contains an overall frame for transformation of knowledge and experience from risk analysis to emergency education. An accident model has been developed to describe the emergency situation. A key concept of this model is uncontrolled flow of energy (UFOE), essential elements are the state, location and movement of the energy (and mass). A UFOE can be considered as the driving force of an accident, e.g., an explosion, a fire, a release of heavy gases. As long as the energy is confined, i.e. the location and movement of the energy are under control, the situation is safe, but loss of confinement will create a hazardous situation that may develop into an accident. A domain model has been developed for representing accident and emergency scenarios occurring in society. The domain model uses three main categories: status, context and objectives. A domain is a group of activities with allied goals and elements and ten specific domains have been investigated: process plant, storage, nuclear power plant, energy distribution, marine transport of goods, marine transport of people, aviation, transport by road, transport by rail and natural disasters. Totally 25 accident cases were consulted and information was extracted for filling into the schematic representations with two to four cases pr. specific domain. (au) 41 tabs., 8 ills.; 79 refs

  3. ACCIDENT PREDICTION METHODOLOGY USING CONFLICT ZONE METHOD FOR “TRANSIT TRANSPORT-PEDESTRIAN” CONFLICT SITUATION AND MODELS OF TRAFFIC FLOWS AT CONTROLLED INTERSECTION

    OpenAIRE

    D. V. Kapsky; P. A. Pegin

    2015-01-01

    Accidents are considered as the most significant cost of road traffic. Therefore any measures for road traffic management should be evaluated according to a minimization  criterion of accident losses. In order to develop a method for evaluation of the accident losses it is necessary to prepare a methodology for cost estimate of road accidents of various severity with due account of their consequences and prediction (economic assessment) and severity level of their consequences (quantitative r...

  4. Reactor physics modelling of accident tolerant fuel for LWRs using ANSWERS codes

    Directory of Open Access Journals (Sweden)

    Lindley Benjamin A.

    2016-01-01

    adopts an integral configuration and a fully passive decay heat removal system to provide indefinite cooling capability for a class of accidents. This paper presents the equilibrium cycle core design and reactor physics behaviour of the I2S-LWR with U3Si2 and the advanced steel cladding. The results were obtained using the traditional two-stage approach, in which homogenized macroscopic cross-section sets were generated by WIMS and applied in a full 3D core solution with PANTHER. The results obtained with WIMS/PANTHER were compared against the Monte Carlo Serpent code developed by VTT and previously reported results for the I2S-LWR. The results were found to be in a good agreement (e.g. <200 pcm in reactivity among the compared codes, giving confidence that the WIMS/PANTHER reactor physics package can be reliably used in modelling advanced LWR systems.

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

    OpenAIRE

    Min Yoo; Sung Min Shin; Hyun Gook Kang

    2014-01-01

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

  6. Radionuclides from the Fukushima accident in the air over Lithuania: measurement and modelling approaches

    International Nuclear Information System (INIS)

    Analyses of 131I, 137Cs and 134Cs 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 131I and 137Cs ranged from 12 μBq/m3 and 1.4 μBq/m3 to 3700 μBq/m3 and 1040 μBq/m3, respectively. The activity concentration of 239,240Pu in one aerosol sample collected from 23 March to 15 April, 2011 was found to be 44.5 nBq/m3. 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. 7Be and 212Pb 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 131I and 137Cs were found to be by 4 orders of magnitude lower as compared to the Chernobyl accident. The activity ratio of 134Cs/137Cs was around 1 with small variations only. The activity ratio of 238Pu/239,240Pu in the aerosol sample was 1.2, indicating a presence of the spent fuel of different origin than that of the Chernobyl accident. - Highlights: ► Two observed maxima in radionuclide concentrations were related to air mass transport. ► HYSPLIT backward trajectories were applied for data interpretation. ► 7Be and 212Pb were used to study a vertical transport of air masses. ► The 134Cs/137Cs activity ratio was around 1. ► 238Pu/239,240Pu ratio was different from global fallout and Chernobyl accident.

  7. Framework for accident management

    International Nuclear Information System (INIS)

    Accident management is an essential element of the Nuclear Regulatory Commission (NRC) Integration Plan for the closure of severe accident issues. This element will consolidate the results from other key elements; such as the Individual Plant Examination (IPE), the Containment Performance Improvement, and the Severe Accident Research Programs, in a form that can be used to enhance the safety programs for nuclear power plants. The NRC is currently conducting an Accident Management Program that is intended to aid in defining the scope and attributes of an accident management program for nuclear power plants. The accident management plan will ensure that a plant specific program is developed and implemented to promote the most effective use of available utility resources (people and hardware) to prevent and mitigate severe accidents. Hardware changes or other plant modifications to reduce the frequency of severe accidents are not a central aim of this program. To accomplish the outlined objectives, the NRC has developed an accident management framework that is comprised of five elements: (1) accident management strategies, (2) training, (3) guidance and computational aids, (4) instrumentation, and (5) delineation of decision making responsibilities. A process for the development of an accident management program has been identified using these NRC framework elements

  8. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    Energy Technology Data Exchange (ETDEWEB)

    Homma, Toshimitsu [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Takahashi, Tomoyuki [Kyoto Univ., Kumatori, Osaka (Japan). Research Reactor Inst; Yonehara, Hidenori [National Inst. of Radiological Sciences, Chiba (Japan)] [eds.

    2000-12-01

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  9. Health effects models for off-site radiological consequence analysis on nuclear reactor accidents (II)

    International Nuclear Information System (INIS)

    This report is a revision of JAERI-M 91-005, 'Health Effects Models for Off-Site Radiological Consequence Analysis of Nuclear Reactor Accidents'. This revision provides a review of two revisions of NUREG/CR-4214 reports by the U.S. Nuclear Regulatory Commission which is the basis of the JAERI health effects models and other several recent reports that may impact the health effects models by international organizations. The major changes to the first version of the JAERI health effects models and the recommended parameters in this report are for late somatic effects. These changes reflect recent changes in cancer risk factors that have come from longer followup and revised dosimetry in major studies on the Japanese A-bomb survivors. This report also provides suggestions about future revisions of computational aspects on health effects models. (author)

  10. Modelling of cladding oxidation by air under severe accident conditions with the MAAP 4 code

    International Nuclear Information System (INIS)

    In a nuclear power plant, air ingress into the vessel is a potential risk in some low probable situations of severe accidents. Air is a highly oxidizing atmosphere that can lead to an enhanced core oxidation and degradation affecting the release of FP. This is particularly true speaking about ruthenium release, which can be significantly increased in the presence of air. This is a key issue due to the high radio-toxicity of ruthenium and its ability to form highly volatile oxides. The oxygen affinity is decreasing in priority from the Zircaloy cladding, to fuel and ruthenium inclusions. It is consequently of great need to understand the phenomena governing cladding oxidation by air as a prerequisite for the source term issues in such scenarios. As a first step, a phenomenological study has been carried out to characterize nitriding of the Zircaloy claddings. In summary, nitriding occurs preferentially when the oxygen has been consumed locally or in case of total oxygen starvation and when the cladding was slightly pre-oxidized. Just like oxidation, nitriding can be modeled in a simplified form as a cladding weight gain in terms of thickness. The model implemented in MAAP takes this into account as well as re-oxidation of the nitrides, in the case where oxygen is available again (especially during a reflood). Several correlations were thus integrated and a new one, called “KIT-EDF”, was developed, based on KIT separate-effect tests. The model has been implemented and validated against QUENCH-16 and QUENCH-10 experiments, studying the oxidation in air atmosphere of an assembly pre-oxidized in steam and finally quenched with water. The simulations give encouraging results since the modeling of nitriding effects has increased hydrogen production during reflood, as experimentally observed. The results of this study lead us to identify a number of perspectives for the future, namely taking into account the changes in the structure of the oxide layer during a

  11. Low-power and shutdown models for the accident sequence precursor (ASP) program

    Energy Technology Data Exchange (ETDEWEB)

    Sattison, M.B.; Thatcher, T.A.; Knudsen, J.K. [Idaho National Engineering Lab., Idaho Falls, ID (United States)] [and others

    1997-02-01

    The US Nuclear Regulatory Commission (NRC) has been using full-power. Level 1, limited-scope risk models for the Accident Sequence Precursor (ASP) program for over fifteen years. These models have evolved and matured over the years, as have probabilistic risk assessment (PRA) and computer technologies. Significant upgrading activities have been undertaken over the past three years, with involvement from the Offices of Nuclear Reactor Regulation (NRR), Analysis and Evaluation of Operational Data (AEOD), and Nuclear Regulatory Research (RES), and several national laboratories. Part of these activities was an RES-sponsored feasibility study investigating the ability to extend the ASP models to include contributors to core damage from events initiated with the reactor at low power or shutdown (LP/SD), both internal events and external events. This paper presents only the LP/SD internal event modeling efforts.

  12. Framework for accident management

    International Nuclear Information System (INIS)

    A program is being conducted to establish those attributes of a severe accident management plan which are necessary to assure effective response to all credible severe accidents and to develop guidance for their incorporation in a plant's Accident Management Plan. This program is one part of the Accident Management Research Program being conducted by the U. S. Nuclear Regulatory Commission (NRC). The approach used in establishing attributes and developing guidance includes three steps. In the first step the general attributes of an accident management plan were identified based on: (1) the objectives established for the NRC accident management program, (2) the elements of an accident management framework identified by the NRC, and (3) a review of the processes used in developing the currently used approach for classifying and analyzing accidents. For the second step, a process was defined that uses the general attributes identified from the first step to develop an accident management plan. The third step applied the process defined in the second step at a nuclear power plant to refine and develop it into a benchmark accident management plan. Step one is completed, step two is underway and step three has not yet begun

  13. A combined M5P tree and hazard-based duration model for predicting urban freeway traffic accident durations.

    Science.gov (United States)

    Lin, Lei; Wang, Qian; Sadek, Adel W

    2016-06-01

    The duration of freeway traffic accidents duration is an important factor, which affects traffic congestion, environmental pollution, and secondary accidents. Among previous studies, the M5P algorithm has been shown to be an effective tool for predicting incident duration. M5P builds a tree-based model, like the traditional classification and regression tree (CART) method, but with multiple linear regression models as its leaves. The problem with M5P for accident duration prediction, however, is that whereas linear regression assumes that the conditional distribution of accident durations is normally distributed, the distribution for a "time-to-an-event" is almost certainly nonsymmetrical. A hazard-based duration model (HBDM) is a better choice for this kind of a "time-to-event" modeling scenario, and given this, HBDMs have been previously applied to analyze and predict traffic accidents duration. Previous research, however, has not yet applied HBDMs for accident duration prediction, in association with clustering or classification of the dataset to minimize data heterogeneity. The current paper proposes a novel approach for accident duration prediction, which improves on the original M5P tree algorithm through the construction of a M5P-HBDM model, in which the leaves of the M5P tree model are HBDMs instead of linear regression models. Such a model offers the advantage of minimizing data heterogeneity through dataset classification, and avoids the need for the incorrect assumption of normality for traffic accident durations. The proposed model was then tested on two freeway accident datasets. For each dataset, the first 500 records were used to train the following three models: (1) an M5P tree; (2) a HBDM; and (3) the proposed M5P-HBDM, and the remainder of data were used for testing. The results show that the proposed M5P-HBDM managed to identify more significant and meaningful variables than either M5P or HBDMs. Moreover, the M5P-HBDM had the lowest overall mean

  14. Dosimetric reconstruction of radiological accident by numerical simulations by means associating an anthropomorphic model and a Monte Carlo computation code

    International Nuclear Information System (INIS)

    After a description of the context of radiological accidents (definition, history, context, exposure types, associated clinic symptoms of irradiation and contamination, medical treatment, return on experience) and a presentation of dose assessment in the case of external exposure (clinic, biological and physical dosimetry), this research thesis describes the principles of numerical reconstruction of a radiological accident, presents some computation codes (Monte Carlo code, MCNPX code) and the SESAME tool, and reports an application to an actual case (an accident which occurred in Equator in April 2009). The next part reports the developments performed to modify the posture of voxelized phantoms and the experimental and numerical validations. The last part reports a feasibility study for the reconstruction of radiological accidents occurring in external radiotherapy. This work is based on a Monte Carlo simulation of a linear accelerator, with the aim of identifying the most relevant parameters to be implemented in SESAME in the case of external radiotherapy

  15. The effect of Chernobyl accident on the development of non malignant diseases

    International Nuclear Information System (INIS)

    The early medical complications of Chernobyl accident include post radiation disease, which were diagnosed in 134 subjects affected by ionizing radiation. 28 persons died during the first 100 days after the event. The increase occurrence of coronary heart disease, endocrine, haematological, dermatological and other diseases were observed after disaster in the contaminated territories. We also discussed the impact of ionizing radiation from Chernobyl accident on pregnancy and congenital defects occurrence. Changes following the Chernobyl accident, as the inhabitants migration from contaminated regions, political and economic conversions, led to depression, anxiety, and even to '' epidemic '' of mental diseases. Increased suicide rate, car accidents, alcohol and drug abuse have been observed in this population. Nowadays vegetative neurosis is more often diagnosed in Ukrainian children. Epidemiological studies were conducted on the ionising radiation effect on the health and on the dose of received radiation after Chernobyl accident face numerous problems as the absence of reliable data regarding diseases in the contaminated territories.(authors)

  16. Investigations of the potential for accident mitigation of the conceptical design for a core retention device developed in SR 209

    International Nuclear Information System (INIS)

    The potential for accident mitigation of the conceptual design for the core retention device developed in SR 209, was estimated by core meltdown and containment analyses. Such a device should prevent late fission product release especially due to failure of the containment by overpressurization (according to release categories 5 and 6 of the German Risk Study, Phase A). Assuming ideal operation of the device, the long term course of a meltdown accident can be influenced in a positive sense including the prevention of overpressurization. However, functional operation of the design seems to be highly uncertain for several reasons, i.g. due to uncertainties in the behaviour of the melt in the retention device. The design does not or only insufficiently cover accident scenarios with high pressure in vessel melt, that means for sequences relevant for their high contribution to core melt frequency. The design may have strong negative impacts on the course of meltdown accidents, LOCA's, constructive details and operation of a plant. Implementation of the measure is expected to need extreme effort, without making sense from the technical point of view. The measure at all seems not to be recommendable for accident mitigation. (orig./HP)

  17. Severe accident simulation at Olkiuoto

    Energy Technology Data Exchange (ETDEWEB)

    Tirkkonen, H.; Saarenpaeae, T. [Teollisuuden Voima Oy (TVO), Olkiluoto (Finland); Cliff Po, L.C. [Micro-Simulation Technology, Montville, NJ (United States)

    1995-09-01

    A personal computer-based simulator was developed for the Olkiluoto nuclear plant in Finland for training in severe accident management. The generic software PCTRAN was expanded to model the plant-specific features of the ABB Atom designed BWR including its containment over-pressure protection and filtered vent systems. Scenarios including core heat-up, hydrogen generation, core melt and vessel penetration were developed in this work. Radiation leakage paths and dose rate distribution are presented graphically for operator use in diagnosis and mitigation of accidents. Operating on an graphically for operator use in diagnosis and mitigation of accidents. Operating on an 486 DX2-66, PCTRAN-TVO achieves a speed about 15 times faster than real-time. A convenient and user-friendly graphic interface allows full interactive control. In this paper a review of the component models and verification runs are presented.

  18. Inversion method of source term in nuclear accident based on Gaussian puff model

    International Nuclear Information System (INIS)

    The inverse problem of source terms information estimation in nuclear accident is important for emergency response. In this study a review of data assimilation applied on atmospheric dispersion is given. For the atmospheric dispersion model is nonlinear and with model errors, ensemble Kalman filter is adopted for data assimilation. The dispersion consequences is described by Gaussian puff model, and the source term emission rate and release height is estimated real-time. To determine the best first guess parameters' value and errors, more than 10 twin experiments have been carried on. The results show that the ensemble Kalman filter can be applied successfully to estimate the source term information when there are one or two unknown parameters, the estimated accuracy is related to first guess value, and is impacted by the standard deviation of perturbation. To reduce the estimation error, first guess value setting to the half to two times of true value is recommended. (author)

  19. Development and first application of a new tool for the simulation of the initiating phase of a severe accident on SFR

    Science.gov (United States)

    Guyot, M.; Gubernatis, P.; Suteau, C.

    2014-06-01

    In order to improve the safety level of Sodium Fast Reactors, low probability events such as Hypothetical Core Disruptive Accident (HCDA) are analyzed for their potential consequences. The initiating phase of such accidents is of particular interest both for the prevention and the mitigation of routes leading to a large core disruption and recriticalities. Up to now, analysis of the initiating phase of HCDA has been performed with the SAS4A code. The SAS4A accident calculations are based on a multiple-channel approach, which requires that subassemblies or groups of similar subassemblies be represented together as independent channels. The SAS4A severe accident calculation scheme resorts to a simplified treatment in which an average pin is used to represent a channel. A point kinetics model coupled with a feedback reactivity model is also used to provide an estimate of the reactor power level. Both to increase the accuracy and decrease the uncertainties in the prediction of reactor safety margins, a new computational tool is currently under development at CEA Cadarache. The main features of this tool are the ability to provide a detailed sub-channel meshing of the sub-assembly as well as three-dimensional kinetics during severe accident conditions. To fulfill these goals, the fluid-dynamics SIMMER-III code has been coupled to the SNATCH solver using a MPI environment. This coupling allows both to compute the multi-phase and multi-component flows encountered in severe accident conditions and to model the power shape variation during voiding and melting of the different reactor materials. This new calculation scheme relies on a SAS-like multiple-channel treatment, where channel-to-channel heat and momentum exchanges are neglected. In this paper, an overview of the SIMMER-III/SNATCH coupled tool capabilities is provided. A first application of this new tool is also performed and compared with a SAS4A reference calculation. The new SIMMER-III/SNATCH tool proved to be

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2014-10-15

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

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

    International Nuclear Information System (INIS)

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

  2. Description of steam flow and structural heat-up during a hypothetic core meltdown accident by a modular computer model

    International Nuclear Information System (INIS)

    The assumption of a reactor core left alone without any cooling facility leads to the hypothetical core meltdown accident. Within the scope of a modular system of computer codes for description by calculation of the relevant heat-up and failure phenomena within the reactor pressure vessel there is established a model for simulating the interaction between partly flooded reactor vessel including internals and dried out region of the active core, this model being applied in the two code units KOCH and UMGEB. By KOCH the time and space dependant steam supply in the core is specified, by UMGEB heating of the support structure relevant for the accident sequence is coupled to the active core region. Both code units together with the other modules of the system allow a flexible simulation of the core meltdown accident. This is demonstrated for a 1000 MWe model reactor by means of the two examples 'flooded core' and 'dry core'. (orig.)

  3. Explaining and predicting workplace accidents using data-mining techniques

    Energy Technology Data Exchange (ETDEWEB)

    Rivas, T., E-mail: trivas@uvigo.e [Dpto. Ingenieria de los Recursos Naturales y Medio Ambiente, E.T.S.I. Minas, University of Vigo, Campus Lagoas, 36310 Vigo (Spain); Paz, M., E-mail: mpaz.minas@gmail.co [Dpto. Ingenieria de los Recursos Naturales y Medio Ambiente, E.T.S.I. Minas, University of Vigo, Campus Lagoas, 36310 Vigo (Spain); Martin, J.E., E-mail: jmartin@cippinternacional.co [CIPP International, S.L. Parque Tecnologico de Asturias, Parcela 43, Oficina 11, 33428 Llanera (Spain); Matias, J.M., E-mail: jmmatias@uvigo.e [Dpto. Estadistica e Investigacion Operativa, E.T.S.I. Minas, University of Vigo, Campus Lagoas, 36310 Vigo (Spain); Garcia, J.F., E-mail: jgarcia@cippinternacional.co [CIPP International, S.L. Parque Tecnologico de Asturias, Parcela 43, Oficina 11, 33428 Llanera (Spain); Taboada, J., E-mail: jtaboada@uvigo.e [Dpto. Ingenieria de los Recursos Naturales y Medio Ambiente, E.T.S.I. Minas, University of Vigo, Campus Lagoas, 36310 Vigo (Spain)

    2011-07-15

    Current research into workplace risk is mainly conducted using conventional descriptive statistics, which, however, fail to properly identify cause-effect relationships and are unable to construct models that could predict accidents. The authors of the present study modelled incidents and accidents in two companies in the mining and construction sectors in order to identify the most important causes of accidents and develop predictive models. Data-mining techniques (decision rules, Bayesian networks, support vector machines and classification trees) were used to model accident and incident data compiled from the mining and construction sectors and obtained in interviews conducted soon after an incident/accident occurred. The results were compared with those for a classical statistical techniques (logistic regression), revealing the superiority of decision rules, classification trees and Bayesian networks in predicting and identifying the factors underlying accidents/incidents.

  4. Explaining and predicting workplace accidents using data-mining techniques

    International Nuclear Information System (INIS)

    Current research into workplace risk is mainly conducted using conventional descriptive statistics, which, however, fail to properly identify cause-effect relationships and are unable to construct models that could predict accidents. The authors of the present study modelled incidents and accidents in two companies in the mining and construction sectors in order to identify the most important causes of accidents and develop predictive models. Data-mining techniques (decision rules, Bayesian networks, support vector machines and classification trees) were used to model accident and incident data compiled from the mining and construction sectors and obtained in interviews conducted soon after an incident/accident occurred. The results were compared with those for a classical statistical techniques (logistic regression), revealing the superiority of decision rules, classification trees and Bayesian networks in predicting and identifying the factors underlying accidents/incidents.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2010-07-01

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

  6. Food safety after nuclear accidents. A Nordic model for national response

    International Nuclear Information System (INIS)

    The Nordic model for the management of food supplies and food safety after nuclear accidents addresses production distribution, sale, and consumption of food and drink. The model contains specific recommendations on intervention levels for distribution and consumption. The overriding aim is to keep the radiation dose to the population as low as reasonably achievable by the optimization of countermeasures. Upper levels of radiation doses which should not be exceeded are termed Primary Intervention Levels. A reasonable maximum dose level resulting from intake of food over a one-year period would be 1 mSv, and this level has been chosen as the starting point for the Nordic model. Maximum levels of radioactive substances in foodstuffs, are termed Derived Intervention Levels (DILs). DILs are established on the basis of the Primary Intervention Levels. A conservative approach is taken which involves additional precautionary assumptions and an extra margin of safety. Provided the DILs are adhered to, the actual radiation dose to which the population is exposed will constitute only a small fraction of the Primary Intervention Levels. The need may arise for specific dietary advise for certain types of food consumed by special population groups. The intervention levels must be adjusted if they cause adverse effects which are unacceptable to the population in general, for instance unfavourable socio-economic impacts. In extreme nuclear accident situations, it may become necessary to suspend the use of intervention levels for a period of time. The full report with scientific annexes was adopted by AeK-LIVS in April 1991, and published as report no. 1991:546 in the Nordic seminar series. In November 1991 the Nordic Council of Ministers requested that the model should be implemented by the national authorities in each of the Nordic countries. The publication contains an abbreviated version of the report. (EG)

  7. Radioactive materials transport accident analysis

    Energy Technology Data Exchange (ETDEWEB)

    McSweeney, T.I.; Maheras, S.J.; Ross, S.B. [Battelle Memorial Inst. (United States)

    2004-07-01

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

  8. Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents. Part 2: IDAC performance influencing factors model

    International Nuclear Information System (INIS)

    This is the second in a series of five papers describing the information, decision, and action in crew context (IDAC) 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, psychological, and physical activities during the course of an accident. This paper identifies the IDAC set of performance influencing factors (PIFs), providing their definitions and causal organization in the form of a modular influence diagram. Fifty PIFs are identified to support the IDAC model to be implemented in a computer simulation environment. They are classified into eleven hierarchically structured groups. The PIFs within each group are independent to each other; however, dependencies may exist between PIFs within different groups. The supporting evidence for the selection and organization of the influence paths based on psychological literature, observations, and various human reliability analysis methodologies is also indicated

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

    International Nuclear Information System (INIS)

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

  10. Review of the chronic exposure pathways models in MACCS [MELCOR Accident Consequence Code System] and several other well-known probabilistic risk assessment models

    International Nuclear Information System (INIS)

    The purpose of this report is to document the results of the work performed by the author in connection with the following task, performed for US Nuclear Regulatory Commission, (USNRC) Office of Nuclear Regulatory Research, Division of Systems Research: MACCS Chronic Exposure Pathway Models: Review the chronic exposure pathway models implemented in the MELCOR Accident Consequence Code System (MACCS) and compare those models to the chronic exposure pathway models implemented in similar codes developed in countries that are members of the OECD. The chronic exposures concerned are via: the terrestrial food pathways, the water pathways, the long-term groundshine pathway, and the inhalation of resuspended radionuclides pathway. The USNRC has indicated during discussions of the task that the major effort should be spent on the terrestrial food pathways. There is one chapter for each of the categories of chronic exposure pathways listed above

  11. Code Development on Aerosol Behavior under Severe Accident-Aerosol Coagulation

    Energy Technology Data Exchange (ETDEWEB)

    Ha, Kwang Soon; Kim, Sung Il; Ryu, Eun Hyun [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    The behaviors of the larger aerosol particles are described usually by continuum mechanics. The smallest particles have diameters less than the mean free path of gas phase molecules and the behavior of these particles can often be described well by free molecular physics. The vast majority of aerosol particles arising in reactor accident analyses have behaviors in the very complicated regime intermediate between the continuum mechanics and free molecular limit. The package includes initial inventories, release from fuel and debris, aerosol dynamics with vapor condensation and revaporization, deposition on structure surfaces, transport through flow paths, and removal by engineered safety features. Aerosol dynamic processes and the condensation and evaporation of fission product vapors after release from fuel are considered within each MELCOR control volume. The aerosol dynamics models are based on MAEROS, a multi-section, multicomponent aerosol dynamics code, but without calculation of condensation. Aerosols can deposit directly on surfaces such as heat structures and water pools, or can agglomerate and eventually fall out once they exceed the largest size specified by the user for the aerosol size distribution. Aerosols deposited on surfaces cannot currently be resuspended.

  12. Modeling operator actions during a small break loss-of-coolant accident in a Babcock and Wilcox nuclear power plant

    International Nuclear Information System (INIS)

    A small break loss-of-accident (SBLOCA) in a typical Babcock and Wilcox (B ampersand W) nuclear power plant was modeled using RELAP5/MOD3. This work was performed as part of the United States Regulatory Commission's (USNRC) Code, Scaling, Applicability and Uncertainty (CSAU) study. The break was initiated by severing one high pressure injection (HPI) line at the cold leg. Thus, the small break was further aggravated by reduced HPI flow. Comparisons between scoping runs with minimal operator action, and full operator action, clearly showed that the operator plays a key role in recovering the plant. Operator actions were modeled based on the emergency operating procedures (EOPs) and the Technical Bases Document for the EOPs. The sequence of operator actions modeled here is only one of several possibilities. Different sequences of operator actions are possible for a given accident because of the subjective decisions the operator must make when determining the status of the plant, hence, which branch of the EOP to follow. To assess the credibility of the modeled operator actions, these actions and results of the simulated accident scenario were presented to operator examiners who are familiar with B ampersand W nuclear power plants. They agreed that, in general, the modeled operator actions conform to the requirements set forth in the EOPs and are therefore plausible. This paper presents the method for modeling the operator actions and discusses the simulated accident scenario from the viewpoint of operator actions

  13. Structural stability of the market model after the Three Mile Island accident

    International Nuclear Information System (INIS)

    This article examines the stability of alpha and beta in the market model resulting from the Three Mile Island accident. The data consist of weekly returns on 70 utility stocks. Both a dummy variable test and the Fisher F statistics are utilized to test for stability. In addition to the individual stocks, the 70 utilities are partioned into two portfolios for the test: nuclear and non-nuclear. The main conclusions are: for the non-nuclear portfolio, no change is observed; for the nuclear portfolio, alpha fell and beta rose (the impact, however, is transitory and insignificant); and the behavior of the residuals suggests that the result is consistent with an efficient market. 17 references, 3 tables

  14. Development of an Anthropomorphous, Remote-Controlled Machine for Handling Radiation Accidents

    International Nuclear Information System (INIS)

    In handling radiation accidents personnel should be kept outside dangerous radiation fields wherever possible. It is obvious that human beings should generally be replaced by machines, which should possess the universal attributes of man. but be insensitive to high radiation activities. The master-slave system appears especially suitable. In this sytem the master is a human being who stays away from the radiation field in a safe room equipped with the necessary information and telecommand devices, and controls and watches the work of the slave from there. The master represents the brain of the machine. The slave is the actual machine and to a certain extent the executive motor part of the human being. It is equipped with limb-like devices, some sense organs such as stereo-television eyes, stereo-microphones, and, by means of power force meters, possesses a certain sense of touch. Further development towards a biped walking machine is outlined. The machine must be given human dimensions as existing atomic plants are made for human operation. (author)

  15. Role of Laws and Regulations For Nuclear Energy Installation in Developing Safety Measures Against Accident

    International Nuclear Information System (INIS)

    The energy industry has been considered as an economic development driver. The fundamental safety policy for nuclear facilities is to protect health and safety of the public and the site personnel against undue risks associated with radiation and radioactive materials resulting from normal operation and abnormal conditions. This policy is implemented, based on the as low as reasonably achievable (ALARA) principle for normal operation and the defense-in-depth principle (prevention of the occurrence of anomalies, prevention of the escalation of anomalies into accidents, and prevention of excessive release of radioactive materials into the environment), through establishment of safety guides and standards. More over the consideration of suitable site selection and safety design, verification by safety evaluation, quality assurance for manufacturing, construction and operation, periodic testing and inspection, confirmation by regulatory bodies, and reflection of experienced troubles to safety countermeasures. Are of these paramount importance concepts are applied variety of nuclear facilities, which is, nuclear reactors, uranium enrichment plants, fuel conversion/fabrication plants, reprocessing plants, radioactive waste management facilities, and so on, considering unique features of each facility.

  16. Temporal-spatial Analysis Model of Traffic Accident Severity Based on Cumulative Logistic Model%基于累积Logistic模型的交通事故严重程度时空分析

    Institute of Scientific and Technical Information of China (English)

    马壮林; 邵春福; 董春娇; 王抢

    2011-01-01

    交通事故的发生具有随机性和偶然性,为尽可能地降低交通事故的伤害程度,根据某高速公路典型事故多发段的交通事故统计资料,以交通事故严重程度为因变量,从时间、道路空间结构和交通运行环境等因素中初步选择12个候选自变量,采用混合逐步选择法分析候选自变量与因变量是否显著相关.采用累积Logistic模型建立交通事故严重程度时空分析模型,并从成比例检验、拟合优度检验和预测准确度检验3个方面对模型进行检验.研究结果表明:事故发生时段、季节因素、发生地点、道路线形、坡度、事故涉及车辆数和日标准交通量与年平均日交通量之比与交通事故严重程度显著相关.%The occurrence of traffic accidents has the randomicity and contingency, this article attempts to make a research on the rules of traffic accidents to minimize the severity of traffic accidents. Firstly, according to the statistical data of a typical accident-prone section, accident severity was selected as the dependent variable, and twelve factors from the three aspects of time, road spatial structure and traffic environment were selected as the candidate independent variables. Then, the combined stepwise method was used to analyze the significant correlation between dependent variable and independent variables, and a temporal-spatial analysis model of traffic accident based on cumulative Logistic model was built. The developed model was tested from three aspects, which are score test for the proportional odds assumption, goodness of fit and predictive accuracy. The results show that seven independent variables, which are accident time, seasonal factors, accident location, road alignment, gradient, the number of vehicles involved in accidents and the ratio of daily traffic to annual average daily traffic, are significantly associated with the dependent variable.

  17. Transportation accidents

    International Nuclear Information System (INIS)

    Predicting the possible consequences of transportation accidents provides a severe challenge to an analyst who must make a judgment of the likely consequences of a release event at an unpredictable time and place. Since it is impractical to try to obtain detailed knowledge of the meteorology and terrain for every potential accident location on a route or to obtain accurate descriptions of population distributions or sensitive property to be protected (data which are more likely to be more readily available when one deals with fixed-site problems), he is constrained to make conservative assumptions in response to a demanding public audience. These conservative assumptions are frequently offset by very small source terms (relative to a fixed site) created when a transport vehicle is involved in an accident. For radioactive materials, which are the principal interest of the authors, only the most elementary models have been used for assessing the consequences of release of these materials in the transportation setting. Risk analysis and environmental impact statements frequently have used the Pasquill-Gifford/gaussian techniques for releases of short duration, which are both simple and easy to apply and require a minimum amount of detailed information. However, after deciding to use such a model, the problem of selecting what specific parameters to use in specific transportation situations still presents itself. Additional complications arise because source terms are not well characterized, release rates can be variable over short and long time periods, and mechanisms by which source aerosols become entrained in air are not always obvious. Some approaches that have been used to address these problems will be reviewed with emphasis on guidelines to avoid the Worst-Case Scenario Syndrome

  18. Ruthenium release modelling in air and steam atmospheres under severe accident conditions using the MAAP4 code

    International Nuclear Information System (INIS)

    Highlights: ► We developed a new modelling of fuel oxidation and ruthenium release in the EDF version of the MAAP4 code. ► We validated this model against some VERCORS experiments. ► Ruthenium release prediction quantitatively and qualitatively well reproduced under air and steam atmospheres. - Abstract: In a nuclear power plant (NPP), a severe accident is a low probability sequence that can lead to core fusion and fission product (FP) release to the environment (source term). For instance during a loss-of-coolant accident, water vaporization and core uncovery can occur due to decay heat. These phenomena enhance core degradation and, subsequently, molten materials can relocate to the lower head of the vessel. Heat exchange between the debris and the vessel may cause its rupture and air ingress. After lower head failure, steam and air entering in the vessel can lead to degradation and oxidation of materials that are still intact in the core. Indeed, Zircaloy-4 cladding oxidation is very exothermic and fuel interaction with the cladding material can decrease its melting temperature by several hundred of Kelvin. FP release can thus be increased, noticeably that of ruthenium under oxidizing conditions. Ruthenium is of particular interest because of its high radio-toxicity due to 103Ru and 106Ru isotopes and its ability to form highly volatile compounds, even at room temperature, such as gaseous ruthenium tetra-oxide (RuO4). It is consequently of great need to understand phenomena governing steam and air oxidation of the fuel and ruthenium release as prerequisites for the source term issues. A review of existing data on these phenomena shows relatively good understanding. In terms of oxygen affinity, the fuel is oxidized before ruthenium, from UO2 to UO2+x. Its oxidation is a rate-controlling surface exchange reaction with the atmosphere, so that the stoichiometric deviation and oxygen partial pressure increase. High temperatures combined with the presence of

  19. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    International Nuclear Information System (INIS)

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result of this study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs

  20. Development of a taxonomy of performance influencing factors for human reliability assessment of accident management tasks and its application

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Jae Whan; Jung, Won Dae; Kang, Dae Il; Ha, Jae Joo

    1999-06-01

    In this study, a new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. We collected the existing PIF taxonomies as many as possible. Then, we analyzed the trend in the selection of PIFs, the frequency of use between PIFs in HRA methods, and the level of definition of PIFs, in order to reflect these characteristics into the development of a new PIF taxonomy. Next, we analyzed the principal task context during accident management to draw the context specific PIFs. Afterwards, we established several criteria for the selection of the appropriate PIFs for HRA under emergency operation and accident management situations. Finally, the final PIF taxonomy containing the subitems for assessing each PIF was constructed based on the results of the previous steps and the selection criteria. The final result ofthis study is the new PIF taxonomy for HRA of the tasks during emergency operation and accident management situations. The selected 11 PIFs in the study are as follows: training and experience, availability and quality of information, status and trend of critical parameters, status of safety system/component, time pressure, working environment features, team cooperation and communication, plant policy and safety culture. (author). 35 refs., 23 tabs.

  1. Road Traffic Accident Analysis of Ajmer City Using Remote Sensing and GIS Technology

    Science.gov (United States)

    Bhalla, P.; Tripathi, S.; Palria, S.

    2014-12-01

    With advancement in technology, new and sophisticated models of vehicle are available and their numbers are increasing day by day. A traffic accident has multi-facet characteristics associated with it. In India 93% of crashes occur due to Human induced factor (wholly or partly). For proper traffic accident analysis use of GIS technology has become an inevitable tool. The traditional accident database is a summary spreadsheet format using codes and mileposts to denote location, type and severity of accidents. Geo-referenced accident database is location-referenced. It incorporates a GIS graphical interface with the accident information to allow for query searches on various accident attributes. Ajmer city, headquarter of Ajmer district, Rajasthan has been selected as the study area. According to Police records, 1531 accidents occur during 2009-2013. Maximum accident occurs in 2009 and the maximum death in 2013. Cars, jeeps, auto, pickup and tempo are mostly responsible for accidents and that the occurrence of accidents is mostly concentrated between 4PM to 10PM. GIS has proved to be a good tool for analyzing multifaceted nature of accidents. While road safety is a critical issue, yet it is handled in an adhoc manner. This Study is a demonstration of application of GIS for developing an efficient database on road accidents taking Ajmer City as a study. If such type of database is developed for other cities, a proper analysis of accidents can be undertaken and suitable management strategies for traffic regulation can be successfully proposed.

  2. PRYMA-TO: A model of radionuclide transfer from air into food stuff. Test with data from the Chernobyl accident

    International Nuclear Information System (INIS)

    This report describes a dynamical model developed in the Environmental Institute of the CIEMAT. Its aims are the calculation of the integrated as well as time-dependent concentrations of ''131l and ''137Cs over time in soils, in forage pasture (or other vegetation species), and in milk and meat. The source contamination is assumed to come from a radioactive cloud confined in the atmospheric mixing layer. Data monitored in different locations the days following the Chernobyl accident have been used. The model was tested against post-Chernobyl data from 13 locations around the world, in the framework of the A4 exercise from the BIOMOVS program (Biospheric Models Validation Studies). The performance of the model is illustrated in 9 scenarios which have been chosen of these 13 because they have more information or they are better described. Default Probability Density Functions for the main parameters used by the model have been obtained by statistical processing of some post-Chernobyl evidence. (Author) 30 refs

  3. PRYMA-TO: A model of radionuclide transfer from air into foodstuff. Test with data from the Chernobyl accident

    International Nuclear Information System (INIS)

    This report describes a dynamical model developed in the Environmental Institute of the CIEMAT. Its aims are the calculation of the integrated as well as time-dependent concentrations of ''131 I and ''137 Cs over time in soils, in forage pasture (or other vegetation species), and in milk and meat. The source contamination is assumed to come from a radioactive cloud confined in the atmospheric mixing layer. Data monitored in different locations the days following the Chernobyl accident have been used. The model was tested against post-Chernobyl data from 13 locations around the world, in the framework of the A4 exercise from the BIOMOVS program (Biospheric Models Validation Studies). The performance of the model is illustrated in 9 scenarios which have been chosen of these 13 because they have more information or they are better described. Default Probability Density Functions for the main parameters used by the model have been obtained by statistical processing of some post-Chernobyl evidence

  4. PRYMA-TO: A model of radionuclide transfer from air into food stuff. Test with data from the Chernobyl accident

    Energy Technology Data Exchange (ETDEWEB)

    Garcia-Olivares, A.; Carrasco, E.; Suanez, A.; Josep, L.

    1994-07-01

    This report describes a dynamical model developed in the Environmental Institute of the CIEMAT. Its aims are the calculation of the integrated as well as time-dependent concentrations of ''131l and ''137Cs over time in soils, in forage pasture (or other vegetation species), and in milk and meat. The source contamination is assumed to come from a radioactive cloud confined in the atmospheric mixing layer. Data monitored in different locations the days following the Chernobyl accident have been used. The model was tested against post-Chernobyl data from 13 locations around the world, in the framework of the A4 exercise from the BIOMOVS program (Biospheric Models Validation Studies). The performance of the model is illustrated in 9 scenarios which have been chosen of these 13 because they have more information or they are better described. Default Probability Density Functions for the main parameters used by the model have been obtained by statistical processing of some post-Chernobyl evidence. (Author) 30 refs.

  5. Review the number of accidents in Tehran over a two-year period and prediction of the number of events based on a time-series model

    Science.gov (United States)

    Teymuri, Ghulam Heidar; Sadeghian, Marzieh; Kangavari, Mehdi; Asghari, Mehdi; Madrese, Elham; Abbasinia, Marzieh; Ahmadnezhad, Iman; Gholizadeh, Yavar

    2013-01-01

    Background: One of the significant dangers that threaten people’s lives is the increased risk of accidents. Annually, more than 1.3 million people die around the world as a result of accidents, and it has been estimated that approximately 300 deaths occur daily due to traffic accidents in the world with more than 50% of that number being people who were not even passengers in the cars. The aim of this study was to examine traffic accidents in Tehran and forecast the number of future accidents using a time-series model. Methods: The study was a cross-sectional study that was conducted in 2011. The sample population was all traffic accidents that caused death and physical injuries in Tehran in 2010 and 2011, as registered in the Tehran Emergency ward. The present study used Minitab 15 software to provide a description of accidents in Tehran for the specified time period as well as those that occurred during April 2012. Results: The results indicated that the average number of daily traffic accidents in Tehran in 2010 was 187 with a standard deviation of 83.6. In 2011, there was an average of 180 daily traffic accidents with a standard deviation of 39.5. One-way analysis of variance indicated that the average number of accidents in the city was different for different months of the year (P < 0.05). Most of the accidents occurred in March, July, August, and September. Thus, more accidents occurred in the summer than in the other seasons. The number of accidents was predicted based on an auto-regressive, moving average (ARMA) for April 2012. The number of accidents displayed a seasonal trend. The prediction of the number of accidents in the city during April of 2012 indicated that a total of 4,459 accidents would occur with mean of 149 accidents per day during these three months. Conclusion: The number of accidents in Tehran displayed a seasonal trend, and the number of accidents was different for different seasons of the year. PMID:26120405

  6. On report of natrium flowing out accident in the high speed breeding reactor 'Monju', the Power Reactor and Nuclear Fuel Development Corporation

    International Nuclear Information System (INIS)

    On December 8th, 1995, a natrium flowing out accident of the 2nd cooling system was occurred in the high speed breeding reactor 'Monju' of the Power Reactor and Nuclear Fuel Development Corporation. The Science and Technology Agency determined to set 'A task force to survey and investigate the natrium flowing out accident of Monju' in the Atomic Energy Safety Bureau to promote at first thorough following its cause at joining some specialists on December 10th, to set it on December 11th. The Atomic Energy Safety Bureau conducted an in-situ inspection and survey after December 20th according to Act 68, Item 1, of Law on Regulation of the Nuclear Raw Materials, the Nuclear Fuels and the Reactor. This report shows results of surveys and investigations till then, points to be reconsidered in the Science and Technology Agency, and some response and improvemental methods on a base of teachings of this accident as well as contents of 'A surveying report of the natrium flowing out accident' dated on February 9th, since considerable understanding of cause elucidation of thermometer breakdown formed this accident and other items. This book contains the following contents as outline of this accident, reason of the accident, protection of enlargement after flowing out, effect of natrium flowing out, response to the outsiders at accident occurring by the Corporation, response to outsiders at accident occurring by the Bureau, and so forth. (G.K.)

  7. Application of Gray Markov SCGM1,1c Model to Prediction of Accidents Deaths in Coal Mining

    OpenAIRE

    Lan, Jian-yi; Zhou, Ying

    2014-01-01

    The prediction of mine accident is the basis of aviation safety assessment and decision making. Gray prediction is suitable for such kinds of system objects with few data, short time, and little fluctuation, and Markov chain theory is just suitable for forecasting stochastic fluctuating dynamic process. Analyzing the coal mine accident human error cause, combining the advantages of both Gray prediction and Markov theory, an amended Gray Markov SCGM1,1c model is proposed. The gray SCGM1,1c mod...

  8. MELCOR analysis of the TMI-2 accident

    Energy Technology Data Exchange (ETDEWEB)

    Boucheron, E.A.

    1990-01-01

    This paper describes the analysis of the Three Mile Island-2 (TMI-2) standard problem that was performed with MELCOR. The MELCOR computer code is being developed by Sandia National Laboratories for the Nuclear Regulatory Commission for the purpose of analyzing severe accident in nuclear power plants. The primary role of MELCOR is to provide realistic predictions of severe accident phenomena and the radiological source team. The analysis of the TMI-2 standard problem allowed for comparison of the model predictions in MELCOR to plant data and to the results of more mechanistic analyses. This exercise was, therefore valuable for verifying and assessing the models in the code. The major trends in the TMI-2 accident are reasonably well predicted with MELCOR, even with its simplified modeling. Comparison of the calculated and measured results is presented and, based on this comparison, conclusions can be drawn concerning the applicability of MELCOR to severe accident analysis. 5 refs., 10 figs., 3 tabs.

  9. Analysis of uncertainties caused by the atmospheric dispersion model in accident consequence assessments with UFOMOD

    International Nuclear Information System (INIS)

    Various techniques available for uncertainty analysis of large computer models are applied, described and selected as most appropriate for analyzing the uncertainty in the predictions of accident consequence assessments. The investigation refers to the atmospheric dispersion and deposition submodel (straight-line Gaussian plume model) of UFOMOD, whose most important input variables and parameters are linked with probability distributions derived from expert judgement. Uncertainty bands show how much variability exists, sensitivity measures determine what causes this variability in consequences. Results are presented as confidence bounds of complementary cumulative frequency distributions (CCFDs) of activity concentrations, organ doses and health effects, partially as a function of distance from the site. In addition the ranked influence of the uncertain parameters on the different consequence types is shown. For the estimation of confidence bounds it was sufficient to choose a model parameter sample size of n (n=59) equal to 1.5 times the number of uncertain model parameters. Different samples or an increase of sample size did not change the 5%-95% - confidence bands. To get statistically stable results of the sensitivity analysis, larger sample sizes are needed (n=100, 200). Random or Latin-hypercube sampling schemes as tools for uncertainty and sensitivity analyses led to comparable results. (orig.)

  10. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident

    International Nuclear Information System (INIS)

    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author)

  11. Sensitivity study of the wet deposition schemes in the modelling of the Fukushima accident.

    Science.gov (United States)

    Quérel, Arnaud; Quélo, Denis; Roustan, Yelva; Mathieu, Anne; Kajino, Mizuo; Sekiyama, Thomas; Adachi, Kouji; Didier, Damien; Igarashi, Yasuhito

    2016-04-01

    The Fukushima-Daiichi release of radioactivity is a relevant event to study the atmospheric dispersion modelling of radionuclides. Actually, the atmospheric deposition onto the ground may be studied through the map of measured Cs-137 established consecutively to the accident. The limits of detection were low enough to make the measurements possible as far as 250km from the nuclear power plant. This large scale deposition has been modelled with the Eulerian model ldX. However, several weeks of emissions in multiple weather conditions make it a real challenge. Besides, these measurements are accumulated deposition of Cs-137 over the whole period and do not inform of deposition mechanisms involved: in-cloud, below-cloud, dry deposition. A comprehensive sensitivity analysis is performed in order to understand wet deposition mechanisms. It has been shown in a previous study (Quérel et al, 2016) that the choice of the wet deposition scheme has a strong impact on the assessment of the deposition patterns. Nevertheless, a "best" scheme could not be highlighted as it depends on the selected criteria: the ranking differs according to the statistical indicators considered (correlation, figure of merit in space and factor 2). A possibility to explain the difficulty to discriminate between several schemes was the uncertainties in the modelling, resulting from the meteorological data for instance. Since the move of the plume is not properly modelled, the deposition processes are applied with an inaccurate activity in the air. In the framework of the SAKURA project, an MRI-IRSN collaboration, new meteorological fields at higher resolution (Sekiyama et al., 2013) were provided and allows to reconsider the previous study. An updated study including these new meteorology data is presented. In addition, a focus on several releases causing deposition in located areas during known period was done. This helps to better understand the mechanisms of deposition involved following the

  12. NRC action plan developed as a result of the TMI-2 accident. Volume 2

    International Nuclear Information System (INIS)

    The Action Plan provides a comprehensive and integrated plan for all actions judged necessary by the Nuclear Regulatory Commission to correct or improve the regulation and operation of nuclear facilities based on the experience from the accident at the Three Mile Island, Unit 2, nuclear facility and the official studies and investigations of the accident. The tables included in this volume list the recommendations from the various organizations and task forces investigating the accident at Three Mile Island. The tables are annotated to provide easy references to the associated parts of the Action Plan in Volume 1. The tables are also annotated to provide a shorthand indication of how the various recommendations are treated in the Action Plan

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-07-15

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

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

    International Nuclear Information System (INIS)

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

  15. Development of MAAP5.0.3 Dose Model for Radiation Environment Effect Analysis

    Energy Technology Data Exchange (ETDEWEB)

    Seo, Mi Ro [KHNP-CRI, Daejeon (Korea, Republic of)

    2015-10-15

    The equipment survivability assessment under the severe accident conditions should be performed. For the environmental conditions such as the pressure and temperature, they can be calculated using MAAP (Modular Accident Analysis Program) code. However, since MAAP itself cannot calculate the radiation DOSE, MAAP5 DOSE model should be developed in order to calculate the DOSE rate during the severe accidents. In this study, we developed the MAAP5 DOSE model for spent fuel pool of OPR1000 type NPP and calculated the DOSE to assess the survivability of the facilities in spent fuel pool and fuel handling region. Until now, there are so many uncertainties in the analysis for radiation effect during the severe accident. However, in terms of the establishment of the severe accident management strategy, quantitative analysis in order to find the general trend for radiation increase during the severe accident is useful. For the radiation environmental effect analysis, the previous studies are mainly focused inside the containment. However, after the Fukushima accident, the severe accident phenomena in the SFP have been the great issues in the nuclear industry including Korea. So, in this study, the dose rate for spent fuel building when the severe accident happens in the SFP is calculated using MAAP5 DOSE. As expected, the dose rate is increased right after the spent fuel is partially uncovered. However, the amount of dose is less significant since the rate of temperature increase is much faster than the rate of dose increase.

  16. Learning lessons from Natech accidents - the eNATECH accident database

    Science.gov (United States)

    Krausmann, Elisabeth; Girgin, Serkan

    2016-04-01

    When natural hazards impact industrial facilities that house or process hazardous materials, fires, explosions and toxic releases can occur. This type of accident is commonly referred to as Natech accident. In order to prevent the recurrence of accidents or to better mitigate their consequences, lessons-learned type studies using available accident data are usually carried out. Through post-accident analysis, conclusions can be drawn on the most common damage and failure modes and hazmat release paths, particularly vulnerable storage and process equipment, and the hazardous materials most commonly involved in these types of accidents. These analyses also lend themselves to identifying technical and organisational risk-reduction measures that require improvement or are missing. Industrial accident databases are commonly used for retrieving sets of Natech accident case histories for further analysis. These databases contain accident data from the open literature, government authorities or in-company sources. The quality of reported information is not uniform and exhibits different levels of detail and accuracy. This is due to the difficulty of finding qualified information sources, especially in situations where accident reporting by the industry or by authorities is not compulsory, e.g. when spill quantities are below the reporting threshold. Data collection has then to rely on voluntary record keeping often by non-experts. The level of detail is particularly non-uniform for Natech accident data depending on whether the consequences of the Natech event were major or minor, and whether comprehensive information was available for reporting. In addition to the reporting bias towards high-consequence events, industrial accident databases frequently lack information on the severity of the triggering natural hazard, as well as on failure modes that led to the hazmat release. This makes it difficult to reconstruct the dynamics of the accident and renders the development of

  17. Monitoring Severe Accidents Using AI Techniques

    International Nuclear Information System (INIS)

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

  18. Emergency Resource Allocation Model for Urban Public Facility Accidents%城市公共设施事故应急资源配置优化模型

    Institute of Scientific and Technical Information of China (English)

    刘近远; 李垚; 闫凯

    2015-01-01

    城市公共设施事故发生时,迅速调动应急资源,尽早展开救援,对促进社会发展和保持社会稳定具有重要意义。本文根据城市公共设施突发事故中资源配置问题的特点,运用多目标规划的思想,建立了应急资源配置的多目标规划模型,并通过实验模拟对模型进行了求解验证。此模型可以成功实现城市公共设施事故应急资源的有效配置,为相关部门的决策提供理论指导。%In the face of urban public facility accidents, mobilizing resources quickly and making them arrive at accident location as soon as possible is of great importance to promote social development and maintain social stability.According to the characteristics of urban public facility accidents and the basic idea of multi-objective programming, a multi-objective programming emergency resource allocation model is proposed and proved to be useful after simulated experiment.This model not only can achieve allocation emergency resources effectively, but also provide theoretical guidance for related departments.

  19. Nuclear criticality safety: general. 3. Tokaimura Criticality Accident: Point Model Stochastic Neutronic Interpretation

    International Nuclear Information System (INIS)

    based on the knowledge of the reactivity insertion. 2. Initiation probability for one neutron P(t). 3. Initiation probability with the neutron source PS (t). Japanese specialists told us that the accident happened during the seventh batch pouring. They estimated the keff before and at the end of this operation: After the sixth batch, K=0.981, and at the end of the seventh batch, K=1.030. When the accident happened (neutron burst), 3 $ was inserted in 15 s, so if we suppose a linear insertion, we have a slope equal to 20 c/s. We may write K(t) = 1 + wt with w = 0.2 β = 0.00160/s. During the accident, there was between 14 and 16 kg of uranium with an enrichment of 18.8%. We have calculated PS(t) and we have taken into account six internal source levels: 1. spontaneous fission: 150 to 170 to 200 n/s; 2. (α, n) reactions and others of this type, and amplification of the internal source during the delayed critical phase: 500 to 1000 to 2000 n/s. In Fig. 2, we can see that the initiation occurred almost surely before 7 s and with a probability close to 0.46 before 2 s with a source of 200 n/s. With a source of 2000 n/s, we have higher initiation probabilities; for example, the initiation occurred almost surely before 2 s and with a probability close to 0.77 before 1 s after the critical time. These results are interesting because they show that a supercritical system does not lead immediately to initiation. One may have short supercritical excursion with no neutron production. The point model approach is useful for gaining a good understanding of what can be the stochastic neutronic contribution for the interpretation of criticality accidents. The results described in this paper may be useful for the interpretation of the time delay between the critical state time and the neutron burst. The thought process we have described may be used in the 'real world', that is, with multigroup or continuous-energy simulations

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

    International Nuclear Information System (INIS)

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

  1. Development of Collision Accident Scenario during Nuclear Spent Fuel Maritime Transportation

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Min; Kang, Hyun Gook [KAIST, Daejeon (Korea, Republic of)

    2015-05-15

    Population density of South Korea is much higher than the other countries, and it is peninsula. Therefore, it is expected that major means of transportation of the spent fuel will be maritime transportation rather than overland transportation. Korea Maritime safety Tribunal (KMST) categorized various maritime accident, see table I. Among them, collision accident is one of the most important and complicated accident from Probabilistic Safety Analysis (PSA) point of view. We will show what will happen if the transportation ship is struck by other ship, how to calculate collision energy and probability of the branches on ship-ship collision with Event Tree Analysis (ETA) method. We selected and re-categorized maritime accident that KMST categorized for ship-ship collision analysis of spent fuel transportation ship. Event tree is constructed and collision energy distribution is derived from statistics and equation. And outer and inner hull fracture probabilities are calculated. If outer hull is broken but inner hull is fine, water will be flooded into the space between outer and inner hull. It will decrease mobility of the ship. If inner hull is fractured, water will be flooded into the ship inside. The ship has compartment structure to resist from foundering. Loss of mobility and compartment damage (ultimately it ends with sink) mechanism need to be analyzed to complete transportation ship collision event tree.

  2. Development of Collision Accident Scenario during Nuclear Spent Fuel Maritime Transportation

    International Nuclear Information System (INIS)

    Population density of South Korea is much higher than the other countries, and it is peninsula. Therefore, it is expected that major means of transportation of the spent fuel will be maritime transportation rather than overland transportation. Korea Maritime safety Tribunal (KMST) categorized various maritime accident, see table I. Among them, collision accident is one of the most important and complicated accident from Probabilistic Safety Analysis (PSA) point of view. We will show what will happen if the transportation ship is struck by other ship, how to calculate collision energy and probability of the branches on ship-ship collision with Event Tree Analysis (ETA) method. We selected and re-categorized maritime accident that KMST categorized for ship-ship collision analysis of spent fuel transportation ship. Event tree is constructed and collision energy distribution is derived from statistics and equation. And outer and inner hull fracture probabilities are calculated. If outer hull is broken but inner hull is fine, water will be flooded into the space between outer and inner hull. It will decrease mobility of the ship. If inner hull is fractured, water will be flooded into the ship inside. The ship has compartment structure to resist from foundering. Loss of mobility and compartment damage (ultimately it ends with sink) mechanism need to be analyzed to complete transportation ship collision event tree

  3. Using meteorological ensembles for atmospheric dispersion modelling of the Fukushima nuclear accident

    Science.gov (United States)

    Périllat, Raphaël; Korsakissok, Irène; Mallet, Vivien; Mathieu, Anne; Sekiyama, Thomas; Didier, Damien; Kajino, Mizuo; Igarashi, Yasuhito; Adachi, Kouji

    2016-04-01

    Dispersion models are used in response to an accidental release of radionuclides of the atmosphere, to infer mitigation actions, and complement field measurements for the assessment of short and long term environmental and sanitary impacts. However, the predictions of these models are subject to important uncertainties, especially due to input data, such as meteorological fields or source term. This is still the case more than four years after the Fukushima disaster (Korsakissok et al., 2012, Girard et al., 2014). In the framework of the SAKURA project, an MRI-IRSN collaboration, a meteorological ensemble of 20 members designed by MRI (Sekiyama et al. 2013) was used with IRSN's atmospheric dispersion models. Another ensemble, retrieved from ECMWF and comprising 50 members, was also used for comparison. The MRI ensemble is 3-hour assimilated, with a 3-kilometers resolution, designed to reduce the meteorological uncertainty in the Fukushima case. The ECMWF is a 24-hour forecast with a coarser grid, representative of the uncertainty of the data available in a crisis context. First, it was necessary to assess the quality of the ensembles for our purpose, to ensure that their spread was representative of the uncertainty of meteorological fields. Using meteorological observations allowed characterizing the ensembles' spread, with tools such as Talagrand diagrams. Then, the uncertainty was propagated through atmospheric dispersion models. The underlying question is whether the output spread is larger than the input spread, that is, whether small uncertainties in meteorological fields can produce large differences in atmospheric dispersion results. Here again, the use of field observations was crucial, in order to characterize the spread of the ensemble of atmospheric dispersion simulations. In the case of the Fukushima accident, gamma dose rates, air activities and deposition data were available. Based on these data, selection criteria for the ensemble members were

  4. Internal dose assessment in radiation accidents

    International Nuclear Information System (INIS)

    Although numerous models have been developed for occupational and medical internal dosimetry, they may not be applicable to an accident situation. Published dose coefficients relate effective dose to intake, but if acute deterministic effects are possible, effective dose is not a useful parameter. Consequently, dose rates to the organs of interest need to be computed from first principles. Standard bioassay methods may be used to assess body contents, but, again, the standard models for bioassay interpretation may not be applicable because of the circumstances of the accident and the prompt initiation of decorporation therapy. Examples of modifications to the standard methodologies include adjustment of biological half-times under therapy, such as in the Goiania accident, and the same effect, complicated by continued input from contaminated wounds, in the Hanford 241Am accident. (author)

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  7. SOURCE 2.0 model development: UO2 thermal properties

    International Nuclear Information System (INIS)

    During analysis of CANDU postulated accidents, the reactor fuel is estimated to experience large temperature variations and to be exposed to a variety of environments from highly oxidized to mildly reducing. The exposure of CANDU fuel to these environments and temperatures may affect fission product releases from the fuel and cause degradation of the fuel thermal properties. The SOURCE 2.0 project is a safety analysis code which will model the necessary mechanisms required to calculate fission product release for a variety of accident scenarios, including large break loss of coolant accidents (LOCAs) with or without emergency core cooling. The goal of the model development is to generate models which are consistent with each other and phenomenologically based, insofar as that is possible given the state of theoretical understanding

  8. Optimum modellings of atmospheric diffusion of radioactive effluents and exposure doses in the accident consequence assessment (Level 3 PSA)

    International Nuclear Information System (INIS)

    Atmospheric diffusion and exposure strongly dependent on the environment were firstly considered in the full spectrum of accident consequence assessment to establish based on Korean conditions. An optimum weather category based on Korean climate and site-specific meteorology of Kori region was established by statistical analysis of measured data for 10 years. And a trajectory model was selected as the optimal one in the ACA by reviewing several existing diffusion models. Following aspects were considered in this selection as availability of meteorological data, ability to treat the change to wind direction, easy applicability of the model, and restriction of CPU time and core memory in current computers. Numerical integration method of our own was selected as the optimal dose assessment tool of external exposure. Unit dose rate was firstly computed with this method as the function of energy level of radionuclide, size of lattice, and distance between source and receptor, and then the results were rearranged as the data library for the rapid access to the ACA run. Dynamic ecosystem modelling has been done in order to estimate the seasonal variation of radioactivity for the assessment of ingestion exposure, considering Korean ingestion behavior, agricultural practice and the transportation. There is a lot of uncertainty in a countermeasure model due to the assumed values of parameters such as fraction of population with different shielding factor and driving speed. A new countermeasure model was developed using the concept of fuzzy set theory, since it provided the mathematical tools which could characterize the uncertainty involved in countermeasure modelling. (Author)

  9. Key factors contributing to accident severity rate in construction industry in Iran: a regression modelling approach.

    Science.gov (United States)

    Soltanzadeh, Ahmad; Mohammadfam, Iraj; Moghimbeigi, Abbas; Ghiasvand, Reza

    2016-03-01

    Construction industry involves the highest risk of occupational accidents and bodily injuries, which range from mild to very severe. The aim of this cross-sectional study was to identify the factors associated with accident severity rate (ASR) in the largest Iranian construction companies based on data about 500 occupational accidents recorded from 2009 to 2013. We also gathered data on safety and health risk management and training systems. Data were analysed using Pearson's chi-squared coefficient and multiple regression analysis. Median ASR (and the interquartile range) was 107.50 (57.24- 381.25). Fourteen of the 24 studied factors stood out as most affecting construction accident severity (psafety and health risk management system to reduce ASR. PMID:27092639

  10. Nuclear accidents

    International Nuclear Information System (INIS)

    On 27 May 1986 the Norwegian government appointed an inter-ministerial committee of senior officials to prepare a report on experiences in connection with the Chernobyl accident. The present second part of the committee's report describes proposals for measures to prevent and deal with similar accidents in the future. The committee's evaluations and proposals are grouped into four main sections: Safety and risk at nuclear power plants; the Norwegian contingency organization for dealing with nuclear accidents; compensation issues; and international cooperation

  11. Bicycle accidents.

    Science.gov (United States)

    Lind, M G; Wollin, S

    1986-01-01

    Information concerning 520 bicycle accidents and their victims was obtained from medical records and the victims' replies to questionnaires. The analyzed aspects included risk of injury, completeness of accident registrations by police and in hospitals, types of injuries and influence of the cyclists' age and sex, alcohol, fatigue, hunger, haste, physical disability, purpose of cycling, wearing of protective helmet and other clothing, type and quality of road surface, site of accident (road junctions, separate cycle paths, etc.) and turning manoeuvres.

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1993-09-01

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

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

    International Nuclear Information System (INIS)

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

  14. Modeling and sensitivity analysis of transport and deposition of radionuclides from the Fukushima Daiichi accident

    Directory of Open Access Journals (Sweden)

    X. Hu

    2014-01-01

    Full Text Available The atmospheric transport and ground deposition of radioactive isotopes 131I and 137Cs during and after the Fukushima Daiichi Nuclear Power Plant (FDNPP accident (March 2011 are investigated using the Weather Research and Forecasting/Chemistry (WRF/Chem model. The aim is to assess the skill of WRF in simulating these processes and the sensitivity of the model's performance to various parameterizations of unresolved physics. The WRF/Chem model is first upgraded by implementing a radioactive decay term into the advection-diffusion solver and adding three parameterizations for dry deposition and two parameterizations for wet deposition. Different microphysics and horizontal turbulent diffusion schemes are then tested for their ability to reproduce observed meteorological conditions. Subsequently, the influence on the simulated transport and deposition of the characteristics of the emission source, including the emission rate, the gas partitioning of 131I and the size distribution of 137Cs, is examined. The results show that the model can predict the wind fields and rainfall realistically. The ground deposition of the radionuclides can also potentially be captured well but it is very sensitive to the emission characterization. It is found that the total deposition is most influenced by the emission rate for both 131I and 137Cs; while it is less sensitive to the dry deposition parameterizations. Moreover, for 131I, the deposition is also sensitive to the microphysics schemes, the horizontal diffusion schemes, gas partitioning and wet deposition parameterizations; while for 137Cs, the deposition is very sensitive to the microphysics schemes and wet deposition parameterizations, and it is also sensitive to the horizontal diffusion schemes and the size distribution.

  15. Irradiated fuel behavior during reactivity initiated accidents in LWR's: Status of research and development studies in France

    International Nuclear Information System (INIS)

    There is much interest in the nuclear industry concerning the ability of training simulators to adequately model severe accident conditions, specifically Anticipated Transient Without Scram (ATWS) events. The Pennsylvania Power and Light Co. has recently installed a new simulator which was provided by S3 Technologies. As part of the licensed operator training program, PP ampersand L provides training on Emergency Operating Procedures (EOPs). Since the ATWS event is challenging from both a computational and operational point of view, the Engineering Department was asked to benchmark the new simulator performance. The purpose of this benchmark was to ensure simulator fidelity with EOP basis calculations which are numerically more rigorous. Once acceptable simulator fidelity had been demonstrated, EOPs were evaluated to ensure they could be implemented by the operators. This paper examines the details of the new simulator response for ATWS events, and exposes the PP ampersand L ATWS procedures to further examination. The simulator benchmark was carried out using the PP ampersand L-developed SABRE code which has been benchmarked against plant data and industry accepted codes. For many ATWS scenarios, the new simulator, which is based upon first principles, provides preditions consistent with SABRE. Reactor power levels, consistent with SABRE results, are significantly higher than predicted by the old simulator, and containment pressurization occurs much more rapidly than previously simulated. Additionally, the new simulated reactor water level, pressure and power are far more responsive to perturbations than predicted by the old simulator. This responsiveness is consistent with SABRE predictions and has helped to define modifications to the ATWS emergency operating procedures. The modified procedures enhance the operators ability to respond to ATWS given the much more realistic reactor model

  16. Health effects model for nuclear power plant accident consequence analysis. Part I. Introduction, integration, and summary. Part II. Scientific basis for health effects models

    Energy Technology Data Exchange (ETDEWEB)

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

    1985-07-01

    Analysis of the radiological health effects of nuclear power plant accidents requires models for predicting early health effects, cancers and benign thyroid nodules, and genetic effects. Since the publication of the Reactor Safety Study, additional information on radiological health effects has become available. This report summarizes the efforts of a program designed to provide revised health effects models for nuclear power plant accident consequence modeling. The new models for early effects address four causes of mortality and nine categories of morbidity. The models for early effects are based upon two parameter Weibull functions. They permit evaluation of the influence of dose protraction and address the issue of variation in radiosensitivity among the population. The piecewise-linear dose-response models used in the Reactor Safety Study to predict cancers and thyroid nodules have been replaced by linear and linear-quadratic models. The new models reflect the most recently reported results of the follow-up of the survivors of the bombings of Hiroshima and Nagasaki and permit analysis of both morbidity and mortality. The new models for genetic effects allow prediction of genetic risks in each of the first five generations after an accident and include information on the relative severity of various classes of genetic effects. The uncertainty in modeloling radiological health risks is addressed by providing central, upper, and lower estimates of risks. An approach is outlined for summarizing the health consequences of nuclear power plant accidents. 298 refs., 9 figs., 49 tabs.

  17. Inverse estimation of source parameters of oceanic radioactivity dispersion models associated with the Fukushima accident

    Directory of Open Access Journals (Sweden)

    Y. Miyazawa

    2013-04-01

    Full Text Available With combined use of the ocean–atmosphere simulation models and field observation data, we evaluate the parameters associated with the total caesium-137 amounts of the direct release into the ocean and atmospheric deposition over the western North Pacific caused by the accident of Fukushima Daiichi nuclear power plant (FNPP that occurred in March 2011. The Green's function approach is adopted for the estimation of two parameters determining the total emission amounts for the period from 12 March to 6 May 2011. It is confirmed that the validity of the estimation depends on the simulation skill near FNPP. The total amount of the direct release is estimated as 5.5–5.9 × 1015 Bq, while that of the atmospheric deposition is estimated as 5.5–9.7 × 1015 Bq, which indicates broader range of the estimate than that of the direct release owing to uncertainty of the dispersion widely spread over the western North Pacific.

  18. Inverse estimation of source parameters of oceanic radioactivity dispersion models associated with the Fukushima accident

    Directory of Open Access Journals (Sweden)

    Y. Miyazawa

    2012-10-01

    Full Text Available With combined use of the ocean-atmosphere simulation models and field observation data, we evaluate the parameters associated with the total caesium-137 amounts of the direct release into the ocean and atmospheric deposition over the Western North Pacific caused by the accident of Fukushima Daiichi nuclear power plant (FNPP that occurred in March 2011. The Green's function approach is adopted for the estimation of two parameters determining the total emission amounts for the period from 12 March to 6 May 2011. It is confirmed that the validity of the estimation depends on the simulation skill near FNPP. The total amount of the direct release is estimated as 5.5–5.9 × 1015 Bq, while that of the atmospheric deposition is estimated as 5.5–9.7 × 1015 Bq, which indicates broader range of the estimate than that of the direct release owing to uncertainty of the dispersion widely spread over the Western North Pacific.

  19. Development of emergency plans for accidents at civil nuclear sites in the United Kingdom

    International Nuclear Information System (INIS)

    This paper is about the management of the off-site consequences of an accident at a UK civil nuclear installation. It describes the measures that would be taken to protect the public in the vicinity of an installation after an emergency had been declared; and it sets out the respective responsibilities of the operator and the various authorities who would be involved in the emergency. (author)

  20. 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

    2004-06-15

    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 2nd step of the 3 year project, and the main researches were focused on developing thermal hydraulic models and improving the auditing code according to the PIRT ranking. The input model for SMART-P has been developed with consistent to the current design status documents. The calculation note has been developed with consistent to the current design status documents. The calculation note has been also prepared 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 matric would be utilized as the basis of code assessment in next year.

  1. Development of evaluation method for economic consequences of severe accident at NPP

    International Nuclear Information System (INIS)

    A nuclear power plant accident has onsite and/or offsite consequences. A common framework into which many of the consequences of an accident may be translated is their economic cost. Though there are some consequence analysis codes, these are limited only to estimate only offsite consequence. In this study, economic losses are estimated for the case of onsite and offsite consequences for the sample PWR plant at Yong-Gwang site. For the estimation of offsite consequence, economic database unique to Korean economic structure is utilized as many as possible. By grouping of various cost components, each cost groups are compared each other. For the detailed estimation of offsite decontamination cost, offsite surface around the plant is divided into five types of surface. This division of surface types is agricultural field, wooded land, roof surface for representing house, asphalt road and water surface. But on the water surface, it is assumed that no decontamination operation is required. The comparison shows that at the severe accident, offsite consequences are more severe than onsite consequences. And in agricultural growing season, the consequence even becomes more severe. It also turned out that the importance and selection of much appropriate criteria will play a major role, because the economic consequences are widely varying depending on how the criteria was chosen

  2. Modeling and analysis of the unprotected loss-of-flow accident in the Clinch River Breeder Reactor

    Energy Technology Data Exchange (ETDEWEB)

    Morris, E.E.; Dunn, F.E.; Simms, R.; Gruber, E.E.

    1985-01-01

    The influence of fission-gas-driven fuel compaction on the energetics resulting from a loss-of-flow accident was estimated with the aid of the SAS3D accident analysis code. The analysis was carried out as part of the Clinch River Breeder Reactor licensing process. The TREAT tests L6, L7, and R8 were analyzed to assist in the modeling of fuel motion and the effects of plenum fission-gas release on coolant and clad dynamics. Special, conservative modeling was introduced to evaluate the effect of fission-gas pressure on the motion of the upper fuel pin segment following disruption. For the nominal sodium-void worth, fission-gas-driven fuel compaction did not adversely affect the outcome of the transient. When uncertainties in the sodium-void worth were considered, however, it was found that if fuel compaction occurs, loss-of-flow driven transient overpower phenomenology could not be precluded.

  3. Management of severe accidents

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Behaviour of high pressure coolant injection system (HPCI) in accident of Fukushima Daiichi Nuclear Power Plant, Unit 3, based on thermodynamic model

    International Nuclear Information System (INIS)

    In order to clarify the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 3 is analyzed from the data open to the public. Phase equilibrium process model was introduced in which the vapor and water are at saturation point in the vessels. The present accident scenario assumes that the high pressure coolant injection system (HPCI) did not worked properly, but the steam in the reactor pressure vessel (RPV) leaked through the turbine of HPCI to the suppression chamber since 12/3/2011 12:35. It is assumed that the Tsunami flooded the torus room where the suppression chamber was placed. Proposed accident scenario agrees with the data of the plant parameters obtained just after the accident. If the present scenario is correct, the behavior that operators in the plant stopped HPCI at 13/3 2:42 did not affect seriously on the RPV rupture. (author)

  6. Severe accident management development program for VVER-1000 and VVER-440/213 based on the westinghouse owners group approach

    International Nuclear Information System (INIS)

    The development of the Westinghouse Owners Group Severe Accident Management Guidelines (WOG SAMG) between 1991 and 1994 was initiated in response to the U.S. Nuclear Regulatory Commission (NRC) requirement for addressing the regulatory severe accident concerns. Hence, the WOG SAMG is designed to interface with other existing procedures at the plant and is used in accident sequences that have progressed to the point where these other procedures are not applicable any longer, i.e. following core damage. The primary purpose of the WOG SAMG is to reach a controlled stable state, which can be declared when fission product releases are controlled, challenges to the confinement fission product boundary have been mitigated, and adequate heat removal is provided to the core and the containment. Although the WOG SAMG is a generic severe accident management guidance developed for use by the entirety of the operating Westinghouse PWR plants, provisions have been made in their development to address specific features of individual plants such as confinement type and the feasibility of reactor cavity flooding. Similarly, the generic SAMG does not address unique plant features and equipment, but rather allows for consideration of plant specific features and strategies. This adaptable approach has led to several SAMG development programs for VVER-1000 and VVER-440 type of power plants, under Westinghouse' s lead. The first of these programs carried out to completion was for Temelin NPP - VVER-1000 - in the first quarter of 2003. Other ongoing programs aim at providing a similar work for VVER-440 design, namely Dukovany, Mochovce and Bohunice NPPs. The challenge of adapting the existing generic WOG material to plants other than PWRs mainly arises for VVER-440 because of important differences in confinement design, making it more vulnerable to ex-vessel phenomena such as hydrogen burn. Also, for both eastern designs, cavity flooding strategy requires special consideration and

  7. 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.

  8. Consequences of Windscale accident (October 1957) and study of the validity of the Sutton's mathematical model of atmospheric diffusion (1960)

    International Nuclear Information System (INIS)

    The reactor accident that happens at the number 1 pile of Windscale in 1957 was followed by a discharge of radioactive products into the atmosphere from the 1.X.1957 at 4.30 PM to the 12.X.1957 at 3.10 PM. On october the 11th it was possible to say that there was no more risk either of external irradiation or inhalation. But in adopting a M.A.C. of 0,1 μcurie of iodine 131 per litre of milk, the Authority had to control the milk delivery till november 23rd on a 500 km2 area. On the other hand, this exceptional accident permit to verify that Sutton's atmospheric diffusion model could give an easy means to foresee, with a sufficient approximation, the consequences of a dispersion of radioactive products into the atmosphere. (author)

  9. The measurement of accident-proneness

    NARCIS (Netherlands)

    As, Sicco van

    2001-01-01

    This paper deals with the measurement of accident-proneness. Accidents seem easy to observe, however accident-proneness is difficult to measure. In this paper I first define the concept of accident-proneness, and I develop an instrument to measure it. The research is mainly executed within chemical

  10. Bow-tie model for offshore drilling blowout accident%海上钻井井喷事故的蝴蝶结模型

    Institute of Scientific and Technical Information of China (English)

    薛鲁宁; 樊建春; 张来斌

    2013-01-01

    海上钻井是一个复杂的动态系统,同时又与一般的过程工业一样,可以划分为清晰的操作步骤和流程,因此安全屏障理论的过程模型非常适合分析海上钻井作业的安全.井喷是海上钻井作业的重要威胁,井喷事故模型对于指导海上钻井安全有重要意义.以安全屏障为基础,利用事故树和事件树分析方法,建立了海上钻井井喷事故的蝴蝶结模型.利用事故树方法分析了井喷事件的原因,通过事件树方法分析了井喷发生后火灾爆炸事故的发展过程.通过将事件原因及事故后果综合为一个模型,操作者可以很直观地了解井喷事故的发生发展过程,进而为其寻找相应的预防及控制措施提供指导.最后将“深水地平线”事故应用于该模型,验证了该模型分析海上钻井井喷事故的有效性.%Offshore drilling is a complex and dynamic system. Meanwhile, like any other process industry, it can also be divided into several independent operating steps and procedures. Therefore, the process model of safety barrier is very suitable to analyze safety of offshore drilling operations. Blowout was an important threat to offshore drilling. A blowout accident model is very meaningful to instruct offshore drilling safety. Based on safety barrier theory, a bow-tie model for offshore drilling blowout was established by utilizing fault tree and event tree methods. Causes of an offshore drilling blowout event were analyzed by fault tree, while development processes of fire & explosion accident after a blowout event were analyzed by event tree. Causes of an offshore drilling blowout event and consequences of fire & explosion accident after the blowout were combined into a single model. This model is very convenient for operators to understand the whole generating and developing process of an offshore drilling blowout accident. Therefore, it can be used as a guide for offshore drilling operators to find relevant

  11. Accidents with sulfuric acid

    Directory of Open Access Journals (Sweden)

    Rajković Miloš B.

    2006-01-01

    Full Text Available Sulfuric acid is an important industrial and strategic raw material, the production of which is developing on all continents, in many factories in the world and with an annual production of over 160 million tons. On the other hand, the production, transport and usage are very dangerous and demand measures of precaution because the consequences could be catastrophic, and not only at the local level where the accident would happen. Accidents that have been publicly recorded during the last eighteen years (from 1988 till the beginning of 2006 are analyzed in this paper. It is very alarming data that, according to all the recorded accidents, over 1.6 million tons of sulfuric acid were exuded. Although water transport is the safest (only 16.38% of the total amount of accidents in that way 98.88% of the total amount of sulfuric acid was exuded into the environment. Human factor was the common factor in all the accidents, whether there was enough control of the production process, of reservoirs or transportation tanks or the transport was done by inadequate (old tanks, or the accidents arose from human factor (inadequate speed, lock of caution etc. The fact is that huge energy, sacrifice and courage were involved in the recovery from accidents where rescue teams and fire brigades showed great courage to prevent real environmental catastrophes and very often they lost their lives during the events. So, the phrase that sulfuric acid is a real "environmental bomb" has become clearer.

  12. Accident and emergency management

    International Nuclear Information System (INIS)

    There is an increasing potential for severe accidents as the industrial development tends towards large, centralised production units. In several industries this has led to the formation of large organisations which are prepared for accidents fighting and for emergency management. The functioning of these organisations critically depends upon efficient decision making and exchange of information. This project is aimed at securing and possibly improving the functionality and efficiency of the accident and emergency management by verifying, demonstrating, and validating the possible use of advanced information technology in the organisations mentioned above. With the nuclear industry in focus the project consists of five main activities: 1) The study and detailed analysis of accident and emergency scenarios based on records from incidents and rills in nuclear installations. 2) Development of a conceptual understanding of accident and emergency management with emphasis on distributed decision making, information flow, and control structure sthat are involved. 3) Development of a general experimental methodology for evaluating the effects of different kinds of decision aids and forms of organisation for emergency management systems with distributed decision making. 4) Development and test of a prototype system for a limited part of an accident and emergency organisation to demonstrate the potential use of computer and communication systems, data-base and knowledge base technology, and applications of expert systems and methods used in artificial intelligence. 5) Production of guidelines for the introduction of advanced information technology in the organisations based on evaluation and validation of the prototype system. (author)

  13. Strategy of severe accident physical modeling in view of recent requirements to safety analysis

    International Nuclear Information System (INIS)

    Nuclear power destiny in various states including Russia is not free from questions. Where there is plenty of non-expensive natural gas or coal in a country, the competition of nuclear power with other power sources is especially intense. Until one considers the economic efficiency or environmental impact of the normally operating plant, the estimate of the proponents favorite choice may be rather optimistic in many cases. As soon as safety aspects of nuclear power are concerned it is necessary to answer very significant questions about the dangers resulting from severe accidents. TMI and, to a greater extent, Chernobyl, demonstrated the other aspect of the severe accident problem. It serves no purpose to dwell upon the inadequate reaction of the population on the radiation problem. It is of little use to try to prove that the health consequences of the Chernobyl or some other radiation accident are substantially overestimated. To make an advance one must substantially reduce the severe accident risk. Besides that is is necessary to give a convincing proof that such a reduction has really been made

  14. Conclusions on severe accident research priorities

    International Nuclear Information System (INIS)

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

  15. Accident Statistics

    Data.gov (United States)

    Department of Homeland Security — Accident statistics available on the Coast Guard’s website by state, year, and one variable to obtain tables and/or graphs. Data from reports has been loaded for...

  16. Investigation of the core melt accident in light water reactors

    International Nuclear Information System (INIS)

    In the thesis the core melt accident, heating up and collapsing of the reactor core were investigated. The most important parameters of influence were found and their effect on the development of the accident were shown. A causal diagram was developed representing the great number of events occurring in the course of the core melt accident as well as their mutual dependences. Models were developed and applied for a detailed description of the collapse process, melting of materials, heat and material transport at flow-off of the melted mass and for taking into account steam blocking in the destroyed core sections. (orig.)

  17. Developing mathematical modelling competence

    DEFF Research Database (Denmark)

    Blomhøj, Morten; Jensen, Tomas Højgaard

    2003-01-01

    In this paper we introduce the concept of mathematical modelling competence, by which we mean being able to carry through a whole mathematical modelling process in a certain context. Analysing the structure of this process, six sub-competences are identified. Mathematical modelling competence...... cannot be reduced to these six sub-competences, but they are necessary elements in the development of mathematical modelling competence. Experience from the development of a modelling course is used to illustrate how the different nature of the sub-competences can be used as a tool for finding...... the balance between different kinds of activities in a particular educational setting. Obstacles of social, cognitive and affective nature for the students' development of mathematical modelling competence are reported and discussed in relation to the sub-competences....

  18. Sports Accidents

    CERN Multimedia

    Kiebel

    1972-01-01

    Le Docteur Kiebel, chirurgien à Genève, est aussi un grand ami de sport et de temps en temps médecin des classes genevoises de ski et également médecin de l'équipe de hockey sur glace de Genève Servette. Il est bien qualifié pour nous parler d'accidents de sport et surtout d'accidents de ski.

  19. Development of the scenario-based training system to reduce hazards and prevent accidents during decommissioning of nuclear facilities

    Energy Technology Data Exchange (ETDEWEB)

    Jeong, KwanSeong; Choi, Jong-Won; Moon, JeiKwon; Choi, ByungSeon; Hyun, Dongjun; Lee, Jonghwan; Kim, IkJune; Kim, GeunHo; Kang, ShinYoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Decommissioning of nuclear facilities has to be accomplished by assuring the safety of workers. 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. In the end, the safety during decommissioning of nuclear facilities will be guaranteed under the principle of ALARA.

  20. Radiation risk models for all solid cancers other than those types of cancer requiring individual assessments after a nuclear accident.

    Science.gov (United States)

    Walsh, Linda; Zhang, Wei

    2016-03-01

    In the assessment of health risks after nuclear accidents, some health consequences require special attention. For example, in their 2013 report on health risk assessment after the Fukushima nuclear accident, the World Health Organisation (WHO) panel of experts considered risks of breast cancer, thyroid cancer and leukaemia. For these specific cancer types, use was made of already published excess relative risk (ERR) and excess absolute risk (EAR) models for radiation-related cancer incidence fitted to the epidemiological data from the Japanese A-bomb Life Span Study (LSS). However, it was also considered important to assess all other types of solid cancer together and the WHO, in their above-mentioned report, stated "No model to calculate the risk for all other solid cancer excluding breast and thyroid cancer risks is available from the LSS data". Applying the LSS models for all solid cancers along with the models for the specific sites means that some cancers have an overlap in the risk evaluations. Thus, calculating the total solid cancer risk plus the breast cancer risk plus the thyroid cancer risk can overestimate the total risk by several per cent. Therefore, the purpose of this paper was to publish the required models for all other solid cancers, i.e. all solid cancers other than those types of cancer requiring special attention after a nuclear accident. The new models presented here have been fitted to the same LSS data set from which the risks provided by the WHO were derived. Although it is known already that the EAR and ERR effect modifications by sex are statistically significant for the outcome "all solid cancer", it is shown here that sex modification is not statistically significant for the outcome "all solid cancer other than thyroid and breast cancer". It is also shown here that the sex-averaged solid cancer risks with and without the sex modification are very similar once breast and thyroid cancers are factored out. Some other notable model

  1. Radiation risk models for all solid cancers other than those types of cancer requiring individual assessments after a nuclear accident.

    Science.gov (United States)

    Walsh, Linda; Zhang, Wei

    2016-03-01

    In the assessment of health risks after nuclear accidents, some health consequences require special attention. For example, in their 2013 report on health risk assessment after the Fukushima nuclear accident, the World Health Organisation (WHO) panel of experts considered risks of breast cancer, thyroid cancer and leukaemia. For these specific cancer types, use was made of already published excess relative risk (ERR) and excess absolute risk (EAR) models for radiation-related cancer incidence fitted to the epidemiological data from the Japanese A-bomb Life Span Study (LSS). However, it was also considered important to assess all other types of solid cancer together and the WHO, in their above-mentioned report, stated "No model to calculate the risk for all other solid cancer excluding breast and thyroid cancer risks is available from the LSS data". Applying the LSS models for all solid cancers along with the models for the specific sites means that some cancers have an overlap in the risk evaluations. Thus, calculating the total solid cancer risk plus the breast cancer risk plus the thyroid cancer risk can overestimate the total risk by several per cent. Therefore, the purpose of this paper was to publish the required models for all other solid cancers, i.e. all solid cancers other than those types of cancer requiring special attention after a nuclear accident. The new models presented here have been fitted to the same LSS data set from which the risks provided by the WHO were derived. Although it is known already that the EAR and ERR effect modifications by sex are statistically significant for the outcome "all solid cancer", it is shown here that sex modification is not statistically significant for the outcome "all solid cancer other than thyroid and breast cancer". It is also shown here that the sex-averaged solid cancer risks with and without the sex modification are very similar once breast and thyroid cancers are factored out. Some other notable model

  2. Road Accident Trends in Africa and Europe

    DEFF Research Database (Denmark)

    Jørgensen, N O

    1997-01-01

    The paper decribes trends and suggests prediction models for accident risks in African and European countries......The paper decribes trends and suggests prediction models for accident risks in African and European countries...

  3. Development of Educational and Training Simulator for Emergency Response to Chinese Nuclear Accidents

    International Nuclear Information System (INIS)

    One of the lessons in the emergency response category is that information on the nuclear power plants of neighboring countries should be organized and the consequence can be assessed. In addition, many reactors have been constructed and are under construction on the eastern coast of China recently. Korea might be directly affected by an accident of Chinese nuclear power plant since Korea is located in the westerly belt. performed with the PCTRAN/CPR-1000 module. The result showed that normal operation and DBA conditions were simulated swiftly with the speed of 16 times faster than real time. Thus, it would be a good source term estimation module for the educational and training simulator

  4. Some aspects of the research and development programmes on the behaviour of containments during severe accidents

    International Nuclear Information System (INIS)

    The R and D programmes relating to the behaviour of containments during severe accidents cover several domains: .leaktightness of the containment: this programme concerns the mechanical resistance of the concretes and the cracking criteria, on the one hand, and the leak rate through the porosities or cracks, on the other; . gaseous releases inside the containment. In addition to the releases of steam and fission products from the primary circuit, the gaseous H20 and C02 releases from the concrete must also be studied: firstly during the corium-concrete interaction, and secondly during the heating of the internal surface of the containment which can be raised to a high temperature on contact with the atmosphere, for example during hydrogen combustion; . the release of fission products during the corium-concrete interactions; . the behaviour of the fission products inside the containment, particularly as regards iodine

  5. Tractor accidents in Swedish traffic.

    Science.gov (United States)

    Pinzke, Stefan; Nilsson, Kerstin; Lundqvist, Peter

    2012-01-01

    The objective of this study is to reach a better understanding of accidents on Swedish roads involving tractors and to suggest ways of preventing them. In an earlier study we analyzed police-reported fatal accidents and accidents that led to physical injuries from 1992 to 2005. During each year of this period, tractors were involved in 128 traffic accidents on average, an average of 7 people were killed, 44 sustained serious injuries, and 143 sustained slight injuries. The number of fatalities in these tractor accidents was about 1.3% of all deaths in traffic accidents in Sweden. Cars were most often involved in the tractor accidents (58%) and 15% were single vehicle accidents. The mean age of the tractor driver involved was 39.8 years and young drivers (15-24 years) were overrepresented (30%). We are now increasing the data collected with the years 2006-2010 in order to study the changes in the number of accidents. Special attention will be given to the younger drivers and to single vehicle accidents. Based on the results we aim to develop suggestions for reducing road accidents, e.g. including measures for making farm vehicles more visible and improvement of the training provided at driving schools. PMID:22317543

  6. 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

    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...

  7. Nuclear fuel cycle facility accident analysis handbook

    International Nuclear Information System (INIS)

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

  8. Nuclear fuel cycle facility accident analysis handbook

    Energy Technology Data Exchange (ETDEWEB)

    Ayer, J E; Clark, A T; Loysen, P; Ballinger, M Y; Mishima, J; Owczarski, P C; Gregory, W S; Nichols, B D

    1988-05-01

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

  9. Development of severe accident evaluation technology (level 2 PSA) for sodium-cooled fast reactors. (1) Overview of evaluation methodology development

    International Nuclear Information System (INIS)

    To consolidate the Probabilistic Safety Assessment (Level 2 PSA) for sodium-cooled fast reactors, the analytical methodologies in the core-material relocation phase and the ex-vessel accident sequence should be established. In addition, a technical basis for Level 2 PSA should also be built up to construct the phenomenological event trees and to determine the branch probabilities in them. In the present study, the analytical methodologies above have been developed with the aim of establishing a severe accident evaluation technology for sodium-cooled fast reactors, and the dominant factors in each phase have been identified through sensitivity analyses. The information obtained from these analyses will be effectively reflected on the technical basis which is indispensable for the construction of phenomenological event trees. (author)

  10. RISK ANALYSIS DEVELOPED MODEL

    Directory of Open Access Journals (Sweden)

    Georgiana Cristina NUKINA

    2012-07-01

    Full Text Available Through Risk analysis developed model deciding whether control measures suitable for implementation. However, the analysis determines whether the benefits of a data control options cost more than the implementation.

  11. Accident management information needs

    International Nuclear Information System (INIS)

    In support of the US Nuclear Regulatory Commission (NRC) Accident Management Research Program, a methodology has been developed for identifying the plant information needs necessary for personnel involved in the management of an accident to diagnose that an accident is in progress, select and implement strategies to prevent or mitigate the accident, and monitor the effectiveness of these strategies. This report describes the methodology and presents an application of this methodology to a Pressurized Water Reactor (PWR) with a large dry containment. A risk-important severe accident sequence for a PWR is used to examine the capability of the existing measurements to supply the necessary information. The method includes an assessment of the effects of the sequence on the measurement availability including the effects of environmental conditions. The information needs and capabilities identified using this approach are also intended to form the basis for more comprehensive information needs assessment performed during the analyses and development of specific strategies for use in accident management prevention and mitigation. 3 refs., 16 figs., 7 tabs

  12. Health effects models for nuclear power plant accident consequence analysis. Part 1, Introduction, integration, and summary: Revision 2

    Energy Technology Data Exchange (ETDEWEB)

    Evans, J.S. [Harvard School of Public Health, Boston, MA (United States); Abrahmson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Inhalation Toxicology Research Inst., Albuquerque, NM (United States); Gilbert, E.S. [Battelle Pacific Northwest Lab., Richland, WA (United States)

    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.

  13. Critical evaluation of the experiments and mathematical models for the determination of fission product release from the spherical fuel elements in cases of core heating accidents in modular HTR's

    International Nuclear Information System (INIS)

    In this work, the thermal behaviour of modular reactors in cases of core heating accidents and the physical phenomena relevant for a release of radioactive materials from HTR fuel elements are explained as far as is necessary for understanding the work. The present mathematical models by which the release of radioactive materials from HTR fuel elements due to diffusion or breaking particles in cases of core heating accidents are also described, examined and evaluated with regard to their applicability to module reactors. The experiments used to verify the mathematical models are also evaluated. The mathematical models are in nearly all cases computer programs, which describe the complicated process of releasing radioactive materials quantitative mathematically. One should point out that these models are constantly being developed further, in line with the increasing amount of knowledge. To conclude the work, proposals are made for improving the certainty of information from experiments and mathematical models to determine the release behaviour of modular reactors. (orig./GL)

  14. Development of Innovative Accident Tolerant High Thermal Conductivity UO2-Diamond Composite Fuel Pellets

    Energy Technology Data Exchange (ETDEWEB)

    Tulenko, James [Univ. of Florida, Gainesville, FL (United States); Subhash, Ghatu [Univ. of Florida, Gainesville, FL (United States)

    2016-01-01

    The University of Florida (UF) evaluated a composite fuel consisting of UO2 powder mixed with diamond micro particles as a candidate as an accident-tolerant fuel (ATF). The research group had previous extensive experience researching with diamond micro particles as an addition to reactor coolant for improved plant thermal performance. The purpose of this research work was to utilize diamond micro particles to develop UO2-Diamond composite fuel pellets with significantly enhanced thermal properties, beyond that already being measured in the previous UF research projects of UO2 – SiC and UO2 – Carbon Nanotube fuel pins. UF is proving with the current research results that the addition of diamond micro particles to UO2 may greatly enhanced the thermal conductivity of the UO2 pellets producing an accident-tolerant fuel. The Beginning of life benefits have been proven and fuel samples are being irradiated in the ATR reactor to confirm that the thermal conductivity improvements are still present under irradiation.

  15. Contribution of mesoscopic modeling for flows prediction in cracked concrete buildings in condition of severe accident

    International Nuclear Information System (INIS)

    This Ph.D. thesis aims at characterising and modeling the mechanical behavior of concrete at the mesoscopic scale. The more general scope of this study is the development of mesoscopic model for concrete; this model is to represent the concrete as a heterogeneous medium, taking into account the difference between aggregate and cement paste respecting the grading curve, the model parameters describe the mechanical and thermal behavior of cement paste and aggregates. We are interested in understanding the concrete behaviour, considered one structure. A program of random granular structure valid in 2D and 3D has been developed. This program is interfaced with the Finite Element code CAST3M in order to compute the numerical simulations. A method for numerical representation of the inclusions of concrete was also developed and validated by projection of the geometry on the shape functions, thus eliminating the problems of meshing that made the representation of all aggregates skeleton almost impossible, particularly in 3D. Firstly, the model is studied in two-dimensional and three-dimensional in order to optimize the geometrical model of the inner structure of concrete in terms of the meshing strategy and the smallest size of the aggregate to be taken into account. The results of the 2D and 3D model are analyzed and compared in the case of uniaxial tension and uniaxial compression. The model used is an isotropic unilateral damage model from Fichant [Fichant et al., 1999]. The model allows to simulate both the macroscopic behavior but also with the local studies of the distribution of crack and crack opening. The model shows interesting results on the transition from diffuse to localized damage and is able to reproduce dilatancy in compression. Finally, the mesoscopic model is applied to three simulations: the calculation of the permeability of cracked concrete; the simulation of the hydration of concrete at early age and finally the scale effect illustrated by bending

  16. Development of an Ingestion Pathway Model for AXAIRQ

    Energy Technology Data Exchange (ETDEWEB)

    Simpkins, A.A.

    1999-01-13

    AXAIRQ is a dose mode code used for prospective accident assessment at the Savannah River Site and is primarily used to show regulatory compliance. For completeness of pathway analysis, an ingestion model, AXINGST, has been developed for use with, and incorporation in, AXAIRQ. Currently available ingestion models were referenced as a basis for AXINGST. AXINGST calculates a conservative ingestion dose following an atmospheric release of radionuclides and includes site specific variables where applicable.

  17. A restructuring proposal based on MELCOR for severe accident analysis code development

    Energy Technology Data Exchange (ETDEWEB)

    Park, Sun Hee; Song, Y. M.; Kim, D. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

    2000-03-01

    In order to develop a template based on existing MELCOR code, current data saving and transferring methods used in MELCOR are addressed first. Then a naming convention for the constructed module is suggested and an automatic program to convert old variables into new derived type variables has been developed. Finally, a restructured module for the SPR package has been developed to be applied to MELCOR. The current MELCOR code ensures a fixed-size storage for four different data types, and manages the variable-sized data within the storage limit by storing the data on the stacked packages. It uses pointer to identify the variables between the packages. This technique causes a difficult grasping of the meaning of the variables as well as memory waste. New features of FORTRAN90, however, make it possible to allocate the storage dynamically, and to use the user-defined data type which lead to a restructured module development for the SPR package. An efficient memory treatment and as easy understanding of the code are allowed in this developed module. The validation of the template has been done by comparing the results of the modified code with those from the existing code, and it is confirmed that the results are the same. The template for the SPR package suggested in this report hints the extension of the template to the entire code. It is expected that the template will accelerate the code domestication thanks to direct understanding of each variable and easy implementation of modified or newly developed models. 3 refs., 15 figs., 16 tabs. (Author)

  18. Reference Model Development

    Energy Technology Data Exchange (ETDEWEB)

    Jepsen, Richard [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

    2011-11-02

    Presentation from the 2011 Water Peer Review in which principal investigator discusses project progress to develop a representative set of Reference Models (RM) for the MHK industry to develop baseline cost of energy (COE) and evaluate key cost component/system reduction pathways.

  19. Authority structure and industrial accidents

    OpenAIRE

    As, Sicco van

    2001-01-01

    This paper deals with the influence of organizational characteristics on safety. Accidents are actually caused by individual mistakes. However the underlying causes of accidents are often organizational. The general hypothesis is that the authority structure is a main cause of accident-proneness within organizations. On one side, the most obvious model for a safe organization would be the ideal-typical bureaucracy. On the other side, potential problems are little flexibility and control is ba...

  20. A probabilistic dispersion model applied to the long-range transport of radionuclides from the Chernobyl accident

    DEFF Research Database (Denmark)

    Lauritzen, B.; Mikkelsen, T.

    1999-01-01

    Long-range atmospheric transport of radionuclides from the Chernobyl accident is modelled as an Eulerian diffusion process. From observations of the gross deposition pattern of particulate radiocaesium an effective long-range Eddy diffusivity K of the order of 10(6) m(2) s(-1) is inferred....... A corresponding effective deposition length for caesium, R-Cs, defined las the effective distance from Chernobyl to where the aerosols have been deposited, is found to be R-Cs approximate to 1000 km. From the observations of the regional variability of the Chernobyl fallout a simple probabilistic assessment...

  1. Methods and data for HTGR fuel performance and radionuclide release modeling during normal operation and accidents for safety analysis

    International Nuclear Information System (INIS)

    The previous status report released in 1987 on reference data and calculation models for fission product transport in High-Temperature, Gas-Cooled Reactor (HTGR) safety analyses has been updated to reflect the current state of knowledge in the German HTGR program. The content of the status report has been expanded to include information from other national programs in HTGRs to provide comparative information on methods of analysis and the underlying database for fuel performance and fission product transport. The release and transport of fission products during normal operating conditions and during the accident scenarios of core heatup, water and air ingress, and depressurization are discussed. (orig.)

  2. Radiological assessment by compartment model POSEIDON-R of radioactivity released in the ocean following Fukushima Daiichi accident

    Science.gov (United States)

    Bezhenar, Roman; Maderich, Vladimir; Heling, Rudie; Jung, Kyung Tae; Myoung, Jung-Goo

    2013-04-01

    The modified compartment model POSEIDON-R (Lepicard et al, 2004), was applied to the North-Western Pacific and adjacent seas. It is for the first time, that a compartment model was used in this region, where 25 Nuclear Power Plants (NPP) are operated. The aim of this study is to perform a radiological assessment of the releases of radioactivity due to the Fukushima Daiichi accident. The model predicts the dispersion of radioactivity in water column and in the sediments, and the transfer of radionuclides throughout the marine food web, and the subsequent doses to the population due to the consumption of fishery products. A generic predictive dynamical food-chain model is used instead of concentration factor (CF) approach. The radionuclide uptake model for fish has as central feature the accumulation of radionuclides in the target tissue. Three layer structure of the water column makes it possible to describe deep-water transport adequately. In total 175 boxes cover the Northwestern Pacific, the East China Sea, and the Yellow Sea and East/Japan Sea. Water fluxes between boxes were calculated by averaging three-dimensional currents obtained by hydrodynamic model ROMS over a 10-years period. Tidal mixing between boxes was parameterized. The model was validated on observation data on the Cs-137 in water for the period 1945-2004. The source terms from nuclear weapon tests are regional source term from the bomb tests on Atoll Enewetak and Atoll Bikini and global deposition from weapons tests. The correlation coefficient between predicted and observed concentrations of Cs-137 in the surface water is 0.925 and RMSE=1.43 Bq/m3. A local-scale coastal box was used according POSEIDON's methodology to describe local processes of activity transport, deposition and food web around the Fukushima Daiichi NPP. The source term to the ocean from the Fukushima accident includes a 10-days release of Cs-134 (5 PBq) and Cs-137 (4 PBq) directly into the ocean and 6 and 5 PBq of Cs-134 and

  3. Health effects models for nuclear power plant accident consequence analysis. Modification of models resulting from addition of effects of exposure to alpha-emitting radionuclides: Revision 1, Part 2, Scientific bases for health effects models, Addendum 2

    Energy Technology Data Exchange (ETDEWEB)

    Abrahamson, S. [Wisconsin Univ., Madison, WI (United States); Bender, M.A. [Brookhaven National Lab., Upton, NY (United States); Boecker, B.B.; Scott, B.R. [Lovelace Biomedical and Environmental Research Inst., Albuquerque, NM (United States). Inhalation Toxicology Research Inst.; Gilbert, E.S. [Pacific Northwest Lab., Richland, WA (United States)

    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 Analysis{close_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.

  4. A RISK MEASUREMENT AND MANAGEMENT MODEL FOR PREVENTING UNMANNED AIR VEHICLE ACCIDENTS

    Directory of Open Access Journals (Sweden)

    Hüdayim BAŞAK

    2008-01-01

    Full Text Available In this study, it is aimed to investigate operationally risky areas by analyzing dangers which can arise during the maintenance and flight activities of Unmanned Air Vehicles (UAVs. For this purpose, a risk analysis methodology was introduced and then within the framework of the application, a sample of risk management model was developed. During the development of the model, personal experiences in the area of UAVs were benefited and a risk management technique consisting of five steps used by pioneering international aviation companies in fliht safety was utilized.

  5. The use of Grey System Theory in predicting the road traffic accident in Fars province in Iran

    Directory of Open Access Journals (Sweden)

    Ali Mohammadi

    2011-10-01

    Full Text Available Traffic accidents have become a more and more important factor that restrict the development of economy and threaten the safety of human beings. Considering the complexity and uncertainty of the influencing factors on traffic accidents, traffic accident forecasting can be regarded as a grey system with unknown and known information, so be analyzed by grey system theory. Grey models require only a limited amount of data to estimate the behavior of unknown systems. In this paper, first, the original predicted values of road traffic accidents are separately obtained by the GM (1,1 model, the Verhulst model and the DGM(2,1 model. The results of these models on predicting road traffic accident show that the forecasting accuracy of the GM(1,1 is higher than the Verhulst model and the DGM(2,1 model. Then, the GM(1,1 model is applied to predict road traffic accident in Fars province.

  6. Simulation analysis of environmental risk accident and management of high-sulfur gas field development in complex terrain

    Institute of Scientific and Technical Information of China (English)

    Xiao WANG; Fanghua HAO; Xuan ZHANG; Wen SUN; Hongguang CHENG

    2008-01-01

    Environmental risk of high sulfur gas field exploitation has become one of the hot spots of envir-onmental management studies. Severe gas H2S blowout accidents in recent years have shown that poor under-standing and estimates of the poisonous gas movement could lead to dangerous evacuation delays. It is important to evaluate the real concentration of H2S, especially in complex terrain. Traditional experiential models are not valid in the case of rough terrain, especially in low-lying areas where the gas accumulates. This study, using high sulfur content gas field of Sichuan "Pu Guang gas field" as study object and adopting objective diagnosis of wind field of land following coordinate three dimensions, applied Lagrangian Puff Model and breaking up tech-nique of puffs to simulate the H2S diffusion condition of blowout accidents produced in the high sulfur content gas field of complex terrain area. The results showed that the H2S distribution did not occur mainly in low wind dir-ection, and due to the obstruction of the mountain's body, it accumulated in front of mountain on produced turn over, flowed around submitted jumping type distribution. The mountain waist near the hilltop and low hollow river valley site rapture points simulating contrast showed that the higher the rapture point, the better the diffusing con-dition of pollutant, the distribution of risk sensitive point decided piping rupture environmental risk size combining the H2S diffusion result and residential area dispersing in the study area, synthetic judge located in the high rapture point environmental risk was smaller than the low hollow point, thus it was suggested to carryout laying of lining build of equal high line of higher terrain. According to simulation results, the environmental risk management measures aimed at putting down adverse effects were worked out.

  7. Licensing topical report: the measurement and modelling of time-dependent fission product release from failed HTGR fuel particles under accident conditions

    International Nuclear Information System (INIS)

    The release of fission products from failed fuel particles was measured under simulated accident (core heatup) conditions. A generic model and specific model parameters that describe delayed fission product release from the kernels of failed HTGR fuel particles were developed from the experimental results. The release of fission products was measured from laser-failed BISO ThO2 and highly enriched (HEU) TRISO UC2 particles that had been irradiated to a range of kernel burnups. The burnups were 0.25, 1.4, and 15.7% FIMA for ThO2 particles and 23.5 and 74% FIMA for UC2 particles. The fission products measured were nuclides of xenon, iodine, krypton, tellurium, and cesium

  8. A balance procedure for calculating the model fuel assemblies reflooding during design basis accident and its verification on PARAMETER test facility

    Science.gov (United States)

    Bazyuk, S. S.; Ignat'ev, D. N.; Parshin, N. Ya.; Popov, E. B.; Soldatkin, D. M.; Kuzma-Kichta, Yu. A.

    2013-05-01

    A balance procedure is proposed for estimating the main parameters characterizing the process of model fuel assemblies reflooding of a VVER reactor made on different scales under the conditions of a design basis accident by subjecting them to bottom reflooding1. The proposed procedure satisfactorily describes the experimental data obtained on PARAMETER test facility in the temperature range up to 1200°C. The times of fuel assemblies quenching by bottom reflooding calculated using the proposed procedure are in satisfactory agreement with the experimental data obtained on model fuel assemblies of VVER- and PWR-type reactors and can be used in developing measures aimed at enhancing the safety of nuclear power stations.

  9. Analysis of a 1:4-scale prestressed concrete containment vessel model for severe accident thermal and pressure loading

    International Nuclear Information System (INIS)

    The Nuclear Power Engineering Corporation (NUPEC) of Japan and the U.S. Nuclear Regulatory Commission (NRC), Office of Nuclear Regulatory Research, co-sponsored and jointly funded a Cooperative Containment Research Program at Sandia National Laboratories (SNL) in Albuquerque, New Mexico. As a part of the program, a prestressed concrete containment vessel (PCCV) model was subjected to a series of overpressurization tests at SNL beginning in July 2000 and culminating in a functional failure mode or Limit State Test (LST) in September 2000 and a Structural Failure Mode Test (SFMT) in November 2001. The PCCV model, uniformly scaled at 1:4, is representative of the containment structure of an actual Pressurized Water Reactor (PWR) plant (OHI-3) in Japan. The objectives of the tests were to obtain measurement data for the structural response of the model to pressure loading beyond design basis accident in order to validate analytical modeling, to find pressure capacity of the model, and to observe its failure mechanisms. Following completion of these tests, OECD/NEA/CSNI agreed to conduct an International Standard Problem (ISP) No.48 on Containment Capacity with the goal of comparing the analytical models and results of a group of international organizations with the test results. One aspect of this International Standard Problem was to evaluate the response of the PCCV model to severe accident loadings which included the thermal environment in addition to the pressure load. This paper describes the analyses to calculate the transient and steady state thermal response of the model and the mechanical response to the combined thermal/pressure loading. Differences with the pressure-only response are discussed and the implications for the capacity of full-scale containment vessels are identified. (authors)

  10. Insights of probabilistic risk analysis on the development of severe accident management guidance: a case study for a plant similar to ANGRA I

    International Nuclear Information System (INIS)

    Probabilistic Risk Analysis (PRA) surges as a way to evaluate the risk of Nuclear Power Plants (NPPs) and to quantify it. Its objective was to track sequences of accidents and define mitigate actions to prevent core damage. But when the core is damaged the question is how to avoid releases of radionuclides to the environment. PRA evaluates this scenario too and is input to the Severe Accident Management Guidance (SAMG). This paper aims in the interaction between PRA and SAMG, both under development for the Brazilian NPPs, focusing in one specific Plant Damage State (PDS). The objective is to develop an Accident Progression Event Tree (APET) proposing the mitigate actions for the event, helping to understand the phenomena.(author)

  11. Development of a severe accident module of a nuclear power plant based in the MELCOR nuclear code and its incorporation to the room simulator

    International Nuclear Information System (INIS)

    This work describes the development of the Severe Accidents Module (MAS) based on the Code MELCOR and its incorporation to the Simulator of Classroom of the Group of Nuclear Engineering of the Engineering Faculty (GrINFI) of the National Autonomous University of Mexico (UNAM). The module of Severe Accidents has the purpose of counting with installed and operational capacity for the simulation of accident sequences with capacitation purposes, training, analysis and design. A shallow description of SimAula is presented, and the philosophy used to obtain the interactive version of MELCOR are discussed, as well as its implementation in the atmosphere of SimAula. Finally, after confirming the correct operation of the development of the tool, some possible topics are discussed for specific applications of the MAS. (Author)

  12. Accident prevention in SME using ORM

    DEFF Research Database (Denmark)

    Jørgensen, Kirsten; Duijm, Nijs Jan; Troen, Hanne

    2008-01-01

    Risk perception in SMEs is normally low, and this is closely related to the fact that the chance of a mall enterprise experiencing a serious accident is very small compared to companies that employ a large workforce. This is a fact even though the SMEs together have a higher accident frequency...... compared with large enterprises. To reach the SMEs we must find a way of supporting them, because they normally have neither the time nor the resources to acquire the knowledge and awareness necessary for working with their own safety. The Occupational Risk Model (ORM) developed by the Dutch Workgroup...

  13. Alternative evacuation strategies for nuclear power accidents

    International Nuclear Information System (INIS)

    In the U.S., current protective-action strategies to safeguard the public following a nuclear power accident have remained largely unchanged since their implementation in the early 1980s. In the past thirty years, new technologies have been introduced, allowing faster computations, better modeling of predicted radiological consequences, and improved accident mapping using geographic information systems (GIS). Utilizing these new technologies, we evaluate the efficacy of alternative strategies, called adaptive protective action zones (APAZs), that use site-specific and event-specific data to dynamically determine evacuation boundaries with simple heuristics in order to better inform protective action decisions (rather than relying on pre-event regulatory bright lines). Several candidate APAZs were developed and then compared to the Nuclear Regulatory Commission’s keyhole evacuation strategy (and full evacuation of the emergency planning zone). Two of the APAZs were better on average than existing NRC strategies at reducing either the radiological exposure, the population evacuated, or both. These APAZs are especially effective for larger radioactive plumes and at high population sites; one of them is better at reducing radiation exposure, while the other is better at reducing the size of the population evacuated. - Highlights: • Developed framework to compare nuclear power accident evacuation strategies. • Evacuation strategies were compared on basis of radiological and evacuation risk. • Current strategies are adequate for smaller scale nuclear power accidents. • New strategies reduced radiation exposure and evacuation size for larger accidents

  14. Examining the nonparametric effect of drivers' age in rear-end accidents through an additive logistic regression model.

    Science.gov (United States)

    Ma, Lu; Yan, Xuedong

    2014-06-01

    This study seeks to inspect the nonparametric characteristics connecting the age of the driver to the relative risk of being an at-fault vehicle, in order to discover a more precise and smooth pattern of age impact, which has commonly been neglected in past studies. Records of drivers in two-vehicle rear-end collisions are selected from the general estimates system (GES) 2011 dataset. These extracted observations in fact constitute inherently matched driver pairs under certain matching variables including weather conditions, pavement conditions and road geometry design characteristics that are shared by pairs of drivers in rear-end accidents. The introduced data structure is able to guarantee that the variance of the response variable will not depend on the matching variables and hence provides a high power of statistical modeling. The estimation results exhibit a smooth cubic spline function for examining the nonlinear relationship between the age of the driver and the log odds of being at fault in a rear-end accident. The results are presented with respect to the main effect of age, the interaction effect between age and sex, and the effects of age under different scenarios of pre-crash actions by the leading vehicle. Compared to the conventional specification in which age is categorized into several predefined groups, the proposed method is more flexible and able to produce quantitatively explicit results. First, it confirms the U-shaped pattern of the age effect, and further shows that the risks of young and old drivers change rapidly with age. Second, the interaction effects between age and sex show that female and male drivers behave differently in rear-end accidents. Third, it is found that the pattern of age impact varies according to the type of pre-crash actions exhibited by the leading vehicle. PMID:24642249

  15. CANSWEL-2: a computer model of the creep deformation of Zircaloy cladding under loss-of-coolant accident conditions

    International Nuclear Information System (INIS)

    The CANSWEL-2 code models cladding creep deformation under conditions relevant to a loss-of-coolant accident (LOCA) in a pressurised water reactor (PWR). It considers in detail the centre rod of a 3 x 3 nominally square array, taking into account azimuthal non-uniformities in cladding thickness and temperature, and the mechanical restraint imposed on contact with neighbouring rods. Any of the rods in the array may assume a non-circular shape. Models are included for primary and secondary creep, dynamic phase change and superplasticity when both alpha- and beta-phase Zircaloy are present. A simple treatment of oxidation strengthening is incorporated. Account is taken of the anisotropic creep behaviour of alpha-phase Zircaloy which leads to cladding bowing. The CANSWEL-2 model is used both as a stand-alone code and also as part of the LOCA analysis code MABEL-2. (author)

  16. Understanding of Iodine Behavior Model Under Severe Accident Conditions in NPPs (I)

    International Nuclear Information System (INIS)

    Support from ISAAC code which is developed in Korea will be needed when a regulatory body (KINS) reviews Wolsong specific SAMG submitted by KHNP in late 2009. The SA research area can be divided into thermal hydraulic (T/H) and fission product (FP) areas and international evaluation shows an increasing opinion that past analysis has large uncertainty in the FP area. Using very small manpower (less than 0.5 MY for 3 year period (2007.3-2010.2)), this study reviewed EU results (BIP, ThAI) and ISAAC possibility for experiment simulation. As results, international and domestic technical status until early 2009 was investigated and input model modification based on these investigation was provided. Simulation possibility of EU experiments showed the Benchmark function in MAAP should be activated in ISAAC (by cooperation with MAAP developer, FAI) as a preliminary arrangement and independent model development reflecting those international results could be possible. This documentation will be referred from PHWR analysis for both source term and safety goal and will enhance the internal understanding for the necessity of iodine behavior study under SA conditions. FP (mainly iodine) model development in the next step using this study will contribute to keeping both improved quality and international competitiveness in ISAAC code

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

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2006-01-01

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

  19. ASSESSMENT OF RUNWAY ACCIDENT HAZARDS IN NIGERIA AVIATION SECTOR

    Directory of Open Access Journals (Sweden)

    Akinyemi Olasunkanmi Oriola

    2015-06-01

    Full Text Available Aviation crashes all over the world have recently been on the high rise, stemming from negligence, mechanical faults, weather, ground control errors, pilot errors, taxing and maintenance crew errors as probable reasons for such accidents. This study models the probabilistic risk assessment of runway accident hazards in Nigeria aviation sector. Six categories of runway accident hazards with their corresponding basic events were identified and modeled using fault tree analysis method of probabilistic risk assessment. The six categories of runway accident hazards are runway surface conditions, weather conditions, collision risk, aircraft system failure, approach/takeoff procedures and human factors. The Fault Tree developed is a system of OR-gates and the weights for each hazard were derived through a domain/expert opinion. The estimated probability of occurrence of runway accident which is the top event of the Fault Tree model is 0.2624. Fault Tree Analysis; thus, identifies the most likely root causes of runway accident through importance measures. The results of the analysis show close relationship of runway accidents in Nigeria aviation sector with aircraft system failure, approach/takeoff procedures, human factor, weather conditions and collision risk.

  20. Development of ODS FeCrAl alloys for accident-tolerant fuel cladding

    Energy Technology Data Exchange (ETDEWEB)

    Dryepondt, Sebastien N. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoelzer, David T. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Pint, Bruce A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Unocic, Kinga A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

    2015-09-18

    FeCrAl alloys are prime candidates for accident-tolerant fuel cladding due to their excellent oxidation resistance up to 1400 C and good mechanical properties at intermediate temperature. Former commercial oxide dispersion strengthened (ODS) FeCrAl alloys such as PM2000 exhibit significantly better tensile strength than wrought FeCrAl alloys, which would alloy for the fabrication of a very thin (~250 m) ODS FeCrAl cladding and limit the neutronic penalty from the replacement of Zr-based alloys by Fe-based alloys. Several Fe-12-Cr-5Al ODS alloys where therefore fabricated by ball milling FeCrAl powders with Y2O3 and additional oxides such as TiO2 or ZrO2. The new Fe-12Cr-5Al ODS alloys showed excellent tensile strength up to 800 C but limited ductility. Good oxidation resistance in steam at 1200 and 1400 C was observed except for one ODS FeCrAl alloy containing Ti. Rolling trials were conducted at 300, 600 C and 800 C to simulate the fabrication of thin tube cladding and a plate thickness of ~0.6mm was reached before the formation of multiple edge cracks. Hardness measurements at different stages of the rolling process, before and after annealing for 1h at 1000 C, showed that a thinner plate thickness could likely be achieved by using a multi-step approach combining warm rolling and high temperature annealing. Finally, new Fe-10-12Cr-5.5-6Al-Z gas atomized powders have been purchased to fabricate the second generation of low-Cr ODS FeCrAl alloys. The main goals are to assess the effect of O, C, N and Zr contents on the ODS FeCrAl microstructure and mechanical properties, and to optimize the fabrication process to improve the ductility of the 2nd gen ODS FeCrAl while maintaining good mechanical strength and oxidation resistance.

  1. Deepwater Horizon Accident Investigation Report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2010-09-15

    On the evening of April 20, 2010, a well control event allowed hydrocarbons to escape from the Macondo well onto Transocean's Deepwater Horizon, resulting in explosions and fire on the rig. Eleven people lost their lives, and 17 others were injured. The fire, which was fed by hydrocarbons from the well, continued for 36 hours until the rig sank. Hydrocarbons continued to flow from the reservoir through the wellbore and the blowout preventer (BOP) for 87 days, causing a spill of national significance. BP Exploration and Production Inc. was the lease operator of Mississippi Canyon Block 252, which contains the Macondo well. BP formed an investigation team that was charged with gathering the facts surrounding the accident, analyzing available information to identify possible causes and making recommendations to enable prevention of similar accidents in the future. The BP investigation team began its work immediately in the aftermath of the accident, working independently from other BP spill response activities and organizations. The ability to gather information was limited by a scarcity of physical evidence and restricted access to potentially relevant witnesses. The team had access to partial real-time data from the rig, documents from various aspects of the Macondo well's development and construction, witness interviews and testimony from public hearings. The team used the information that was made available by other companies, including Transocean, Halliburton and Cameron. Over the course of the investigation, the team involved over 50 internal and external specialists from a variety of fields: safety, operations, subsea, drilling, well control, cementing, well flow dynamic modeling, BOP systems and process hazard analysis. This report presents an analysis of the events leading up to the accident, eight key findings related to the causal chain of events and recommendations to enable the prevention of a similar accident. The investigation team worked

  2. Seismic Shaking Table Requirements and Consideration of Fluid-Structure Interaction Effect in Seismic Response Analysis Model for In-Reactor Fuel Assembly Under Severe Earthquake Accident

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Kanghee; Yoon, Kyungho; Kang, Heungsoek; Lee, Youngho; Kim, Hyungkyu [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

    Dynamic response of fuel assembly can be significantly affected by added hydrodynamic mass and additional damping from the fluid and flow inside operating reactor core. Added mass or hydrodynamic virtual mass from surrounding fluid medium can be theoretically estimated by the potential flow theory. Solving Laplace equation in terms of velocity potential can leads to calculate mass components in the mass matrix of simplified fuel FE model. Additional damping from the fluid and the flow inside reactor core are originated from fluid drag and flow lift force, respectively. Lift force from axial flow can increase fuel assembly damping by twice compared to still fluid damping from the loop testing. In practice, fuel assembly damping should be measured by mockup loop testing and referred to published data in the literature. The justification is performed via time history analysis with simplified dynamic model using a group of fuel assembly in the core. Key check points in this analysis might be the integrity of intermediate spacer grids when impacting fuels into core shroud plate or into neighboring fuel assembly. Thus, dynamic displacement and impact force at grid elevations are the important structural parameters to be traced out during the analysis and the simulation testing. KAERI have a plan to develop dynamic analysis model and to setup test infrastructure for full scale and several fuel assembly rows seismic simulation testing. This paper briefly discuss on the reference earthquake accident scenario, shaking table requirements for full-scale seismic simulation testing, virtual testing issues before the hardware setup, and modelling issue related to fluid-structure interaction effect in accident core analysis.

  3. Product Development Process Modeling

    Institute of Scientific and Technical Information of China (English)

    1999-01-01

    The use of Concurrent Engineering and other modern methods of product development and maintenance require that a large number of time-overlapped "processes" be performed by many people. However, successfully describing and optimizing these processes are becoming even more difficult to achieve. The perspective of industrial process theory (the definition of process) and the perspective of process implementation (process transition, accumulation, and inter-operations between processes) are used to survey the method used to build one base model (multi-view) process model.

  4. Estimating radiation doses from reactor accidents

    International Nuclear Information System (INIS)

    In order to plan for emergency response to reactor accidents involving large radiation releases, it is necessary to determine the medical resources, such as diagnostic laboratory tests, hospital facilities and convalescent care, needed to care for a large population exposed to radiation. A determination of the needed medical resources is difficult because of the widely varying sensitivity humans exhibit to radiation exposure and because of the large number of assumptions involved in predicting radiation dispersion. This paper demonstrates a simple method for approximating medical needs in response to a severe reactor accident. The method requires a model for radiation dispersion from the accident and data for population distribution surrounding the reactor. With this information, tables developed in this paper may be used to project medical needs. The needs identified by this methodology may be compared against the actual medical resources of nearby communities to determine the size of the area impacted

  5. Modeling of leachable {sup 137}Cs in throughfall and stemflow for Japanese forest canopies after Fukushima Daiichi Nuclear Power Plant accident

    Energy Technology Data Exchange (ETDEWEB)

    Loffredo, Nicolas, E-mail: wataiso@free.fr [Center for Research in Isotopes and Environmental Dynamics, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8572 (Japan); Onda, Yuichi [Center for Research in Isotopes and Environmental Dynamics, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8572 (Japan); Kawamori, Ayumi [Graduate School of Life and Environmental Sciences, University of Tsukuba (Japan); Kato, Hiroaki [Center for Research in Isotopes and Environmental Dynamics, University of Tsukuba, 1-1-1 Tennodai, Tsukuba, Ibaraki 305-8572 (Japan)

    2014-09-15

    The Fukushima accident dispersed significant amounts of radioactive cesium (Cs) in the landscape. Our research investigated, from June 2011 to November 2013, the mobility of leachable Cs in forests canopies. In particular, {sup 137}Cs and {sup 134}Cs activity concentrations were measured in rainfall, throughfall, and stemflow in broad-leaf and cedar forests in an area located 40 km from the power plant. Leachable {sup 137}Cs loss was modeled by a double exponential (DE) model. This model could not reproduce the variation in activity concentration observed. In order to refine the DE model, the main physical measurable parameters (rainfall intensity, wind velocity, and snowfall occurrence) were assessed, and rainfall was identified as the dominant factor controlling observed variation. A corrective factor was then developed to incorporate rainfall intensity in an improved DE model. With the original DE model, we estimated total {sup 137}Cs loss by leaching from canopies to be 72 ± 4%, 67 ± 4%, and 48 ± 2% of the total plume deposition under mature cedar, young cedar, and broad-leaf forests, respectively. In contrast, with the improved DE model, the total {sup 137}Cs loss by leaching was estimated to be 34 ± 2%, 34 ± 2%, and 16 ± 1% of the total plume deposition under mature cedar, young cedar, and broad-leaf forests, respectively. The improved DE model corresponds better to observed data in literature. Understanding {sup 137}Cs and {sup 134}Cs forest dynamics is important for forecasting future contamination of forest soils around the FDNPP. It also provides a basis for understanding forest transfers in future potential nuclear disasters. - Highlights: • A double exponential model was used to model leachable cesium loss from canopies. • The model could not reproduce variation observed. • Rainfall was identified as the dominant factor controlling the variation. • A rainfall parameter was used to develop an improved double exponential model. • The

  6. Reconstruction of (131)I radioactive contamination in Ukraine caused by the Chernobyl accident using atmospheric transport modelling.

    Science.gov (United States)

    Talerko, Nikolai

    2005-01-01

    The evaluation of (131)I air and ground contamination field formation in the territory of Ukraine was made using the model of atmospheric transport LEDI (Lagrangian-Eulerian DIffusion model). The (131)I atmospheric transport over the territory of Ukraine was simulated during the first 12 days after the accident (from 26 April to 7 May 1986) using real aerological information and rain measurement network data. The airborne (131)I concentration and ground deposition fields were calculated as the database for subsequent thyroid dose reconstruction for inhabitants of radioactive contaminated regions. The small-scale deposition field variability is assessed using data of (137)Cs detailed measurements in the territory of Ukraine. The obtained results are compared with available data of radioiodine daily deposition measurements made at the network of meteorological stations in Ukraine and data of the assessments of (131)I soil contamination obtained from the (129)I measurements. PMID:16024139

  7. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors; Modelisation en dynamique rapide d'accidents dans le circuit primaire des reacteurs a eau pressurisee

    Energy Technology Data Exchange (ETDEWEB)

    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)

  8. Development of a shell finite element. Application to the thermo-viscoplastic behaviour of a PWR vessel during a severe accident; Developpement d`un element fini coque. Application au comportement thermo-viscoplastique d`une cuve de reacteur nucleaire (REP) en situation d`accident grave

    Energy Technology Data Exchange (ETDEWEB)

    Diaz, V

    1998-10-07

    The aim of this study is to develop a model for the thermo-viscoplastic behaviour of he power water reactor lower head during a severe accident, so as to implement it in codes representing the whole accident progress (scenario codes). So it has to give a precise solution in a short cpu-time. The main loadings are the internal pressure and the strong longitudinal and transverse thermal gradients. To deal with this problem, the idea is to develop a new shell element with variable mechanical parameters with the temperature. This is possible in taking advantage of the properties of the bending center line, called neutral fiber. Besides, this new shell element has the particularity to be able to melt without modifying the initial dimensions of the structure. Then, we have developed a complete program to study the mechanical resistance of the vessel. The visco-plastic behaviour is considered as a loading (so it is placed in the second member of the system to be solved) and represented by a Norton law whose parameters depend on the temperature, the law is integrated explicitly which necessitates the introduction of criteria limiting the time step. The rupture criterion by creep is defined by a damage law whereas the rupture criterion by plasticity is based on the exceeding of the mean limit stress in the thickness. Then the model was validated by comparing the results with those of a Castem 2000 volume mesh (finite element code). Finally the model was coupled with the scenario codes ICARE2 and MAAP4 and tested on two typical severe accidents. The results are very satisfactory both on accuracy and cpu-time execution. (author) 113 refs.

  9. FASTGRASS: A mechanistic model for the prediction of Xe, I, Cs, Te, Ba, and Sr release from nuclear fuel under normal and severe-accident conditions

    International Nuclear Information System (INIS)

    The primary physical/chemical models that form the basis of the FASTGRASS mechanistic computer model for calculating fission-product release from nuclear fuel are described. Calculated results are compared with test data and the major mechanisms affecting the transport of fission products during steady-state and accident conditions are identified

  10. Tchernobyl accident

    International Nuclear Information System (INIS)

    First, R.M.B.K type reactors are described. Then, safety problems are dealt with reactor control, behavior during transients, normal loss of power and behavior of the reactor in case of leak. A possible scenario of the accident of Tchernobyl is proposed: events before the explosion, possible initiators, possible scenario and events subsequent to the core meltdown (corium-concrete interaction, interaction with the groundwater table). An estimation of the source term is proposed first from the installation characteristics and the supposed scenario of the accident, and from the measurements in Europe; radiological consequences are also estimated. Radioactivity measurements (Europe, Scandinavia, Western Europe, France) are given in tables (meteorological maps and fallouts in Europe). Finally, a description of the site is given

  11. Accident: Reminder

    CERN Multimedia

    2003-01-01

    There is no left turn to Point 1 from the customs, direction CERN. A terrible accident happened last week on the Route de Meyrin just outside Entrance B because traffic regulations were not respected. You are reminded that when travelling from the customs, direction CERN, turning left to Point 1 is forbidden. Access to Point 1 from the customs is only via entering CERN, going down to the roundabout and coming back up to the traffic lights at Entrance B

  12. The SAM software system for modeling severe accidents at nuclear power plants equipped with VVER reactors on full-scale and analytic training simulators

    Science.gov (United States)

    Osadchaya, D. Yu.; Fuks, R. L.

    2014-04-01

    The architecture of the SAM software package intended for modeling beyond-design-basis accidents at nuclear power plants equipped with VVER reactors evolving into a severe stage with core melting and failure of the reactor pressure vessel is presented. By using the SAM software package it is possible to perform comprehensive modeling of the entire emergency process from the failure initiating event to the stage of severe accident involving meltdown of nuclear fuel, failure of the reactor pressure vessel, and escape of corium onto the concrete basement or into the corium catcher with retention of molten products in it.

  13. Severe accident analysis code Sampson for impact project

    Energy Technology Data Exchange (ETDEWEB)

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

    2001-07-01

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

  14. Investigation of adolescent accident predictive variables in hilly regions.

    Science.gov (United States)

    Mohanty, Malaya; Gupta, Ankit

    2016-09-01

    The study aims to determine the significant personal and environmental factors in predicting the adolescent accidents in the hilly regions taking into account two cities Hamirpur and Dharamshala, which lie at an average elevation of 700--1000 metres above the mean sea level (MSL). Detailed comparisons between the results of 2 cities are also studied. The results are analyzed to provide the list of most significant factors responsible for adolescent accidents. Data were collected from different schools and colleges of the city with the help of a questionnaire survey. Around 690 responses from Hamirpur and 460 responses from Dharamshala were taken for study and analysis. Standard deviations (SD) of various factors affecting accidents were calculated and factors with relatively very low SD were discarded and other variables were considered for correlations. Correlation was developed using Kendall's-tau and chi-square tests and factors those were found significant were used for modelling. They were - the victim's age, the character of road, the speed of vehicle, and the use of helmet for Hamirpur and for Dharamshala, the kind of vehicle involved was an added variable found responsible for adolescent accidents. A logistic regression was performed to know the effect of each category present in a variable on the occurrence of accidents. Though the age and the speed of vehicle were considered to be important factors for accident occurrence according to Indian accident data records, even the use of helmet comes out as a major concern. The age group of 15-18 and 18-21 years were found to be more susceptible to accidents than the higher age groups. Due to the presence of hilly area, the character of road becomes a major concern for cause of accidents and the topography of the area makes the kind of vehicle involved as a major variable for determining the severity of accidents. PMID:26077876

  15. 交通事故宏观计量经济学模型(英文)%Macroscopic econometrics model of traffic accident

    Institute of Scientific and Technical Information of China (English)

    王军雷; 孙小端; 贺玉龙; 侯树展

    2012-01-01

    From the points of macroscopic factors such as economic development level,population number,vehicle ownership and road condition,the regularities of traffic accidents at home and abroad were analyzed.The relations among per capita GDP and vehicle ownership per 1 000 population,mortality per 10 000 vehicles,mortality per 100 000 population were studied.Based on macroscopic econometrics and Cobb-Douglas function,the panel data model of traffic accident was set up by using the historical data in seven countries.Fixed effect model and random effect model were used to estimate parameters respectively,Hausman test was carried out,and the macroscopic econometrics models of traffic accidents for the seven countries were set up.Calculation result shows that among the parameters of traffic accidents,mortality per 100 000 population is negative correlation with per capita GDP and per capita road length,mortality per 100 000 population is positive correlation with vehicle ownership per 1 000 population.Through Hausman test,chi-square distribution value is 3.91 when freedom is 3,the probability is 0.02 and less than the confidence level which is 0.05.Compared with the random effect model,all the confidence levels of variables for fixed effect model are less than 0.05,and the goodness of fit is better.So the fixed effect model is effective.3 tabs,6 figs,18 refs.%从经济发展水平、人口数量、汽车保有量、道路情况等宏观因素入手,研究了国内外道路交通事故规律,分析了人均GDP与千人汽车保有量、万车死亡率、10万人口死亡率之间的关系。以宏观计量经济学和柯布-道格拉斯函数为基础,利用7个国家的历史数据构建了交通事故面板数据模型。分别采用固定效应模型和随机效应模型进行参数估计,并进行了Hausman检验,得到7个国家的交通事故宏观计量经济学模型。计算结果表明:在交通事故参数中,10万人口死亡率与人均GDP、人均道路长度

  16. Behaviour of isolation condensers (ICs) in accident of Fukushima Daiichi Nuclear Power Plant, Unit 1, based on thermodynamic model. And yet, the IC was working after Tsunami arrival

    International Nuclear Information System (INIS)

    In order to investigate the process of Accident of Fukushima Nuclear Plants, an accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 1 is analyzed from the data open to the public. Phase equilibrium process model and adiabatic expansion model were introduced. Original data reported in the first stage of the accident were examined to clarify the behaviour of the isolation condensers (ICs) which are generally believed unfunctional after the arrival of Tsunami and station blackout. The original data and observation reports verified that the so called 'fail safe' system to close the valves in IC did not work properly due to the shutdown of AC power. The reports also showed the evidence that the operators injected water to the storage tank of IC. (author)

  17. IMPACT fragmentation model developments

    Science.gov (United States)

    Sorge, Marlon E.; Mains, Deanna L.

    2016-09-01

    The IMPACT fragmentation model has been used by The Aerospace Corporation for more than 25 years to analyze orbital altitude explosions and hypervelocity collisions. The model is semi-empirical, combining mass, energy and momentum conservation laws with empirically derived relationships for fragment characteristics such as number, mass, area-to-mass ratio, and spreading velocity as well as event energy distribution. Model results are used for several types of analysis including assessment of short-term risks to satellites from orbital altitude fragmentations, prediction of the long-term evolution of the orbital debris environment and forensic assessments of breakup events. A new version of IMPACT, version 6, has been completed and incorporates a number of advancements enabled by a multi-year long effort to characterize more than 11,000 debris fragments from more than three dozen historical on-orbit breakup events. These events involved a wide range of causes, energies, and fragmenting objects. Special focus was placed on the explosion model, as the majority of events examined were explosions. Revisions were made to the mass distribution used for explosion events, increasing the number of smaller fragments generated. The algorithm for modeling upper stage large fragment generation was updated. A momentum conserving asymmetric spreading velocity distribution algorithm was implemented to better represent sub-catastrophic events. An approach was developed for modeling sub-catastrophic explosions, those where the majority of the parent object remains intact, based on estimated event energy. Finally, significant modifications were made to the area-to-mass ratio distribution to incorporate the tendencies of different materials to fragment into different shapes. This ability enabled better matches between the observed area-to-mass ratios and those generated by the model. It also opened up additional possibilities for post-event analysis of breakups. The paper will discuss

  18. Accident management insights after the Fukushima Daiichi NPP accident

    International Nuclear Information System (INIS)

    The Fukushima Daiichi nuclear power plant (NPP) accident, that took place on 11 March 2011, initiated a significant number of activities at the national and international levels to reassess the safety of existing NPPs, evaluate the sufficiency of technical means and administrative measures available for emergency response, and develop recommendations for increasing the robustness of NPPs to withstand extreme external events and beyond design basis accidents. The OECD Nuclear Energy Agency (NEA) is working closely with its member and partner countries to examine the causes of the accident and to identify lessons learnt with a view to the appropriate follow-up actions to be taken by the nuclear safety community. Accident management is a priority area of work for the NEA to address lessons being learnt from the accident at the Fukushima Daiichi NPP following the recommendations of Committee on Nuclear Regulatory Activities (CNRA), Committee on the Safety of Nuclear Installations (CSNI), and Committee on Radiation Protection and Public Health (CRPPH). Considering the importance of these issues, the CNRA authorised the formation of a task group on accident management (TGAM) in June 2012 to review the regulatory framework for accident management following the Fukushima Daiichi NPP accident. The task group was requested to assess the NEA member countries needs and challenges in light of the accident from a regulatory point of view. The general objectives of the TGAM review were to consider: - enhancements of on-site accident management procedures and guidelines based on lessons learnt from the Fukushima Daiichi NPP accident; - decision-making and guiding principles in emergency situations; - guidance for instrumentation, equipment and supplies for addressing long-term aspects of accident management; - guidance and implementation when taking extreme measures for accident management. The report is built on the existing bases for capabilities to respond to design basis

  19. Development of the aerosol generation system for simulating the dry deposition behavior of radioaerosol emitted by the accident of FDNPP

    Science.gov (United States)

    Zhang, Z.

    2015-12-01

    A large amount of radioactivity was discharged by the accident of FDNPP. The long half-life radionuclide, 137Cs was transported through the atmosphere mainly as the aerosol form and deposited to the forests in Fukushima prefecture. After the dry deposition of the 137Cs, the foliar uptake process would occur. To evaluate environmental transfer of radionuclides, the dry deposition and following foliar uptake is very important. There are some pioneering studies for radionuclide foliar uptake with attaching the solution containing stable target element on the leaf, however, cesium oxide aerosols were used for these deposition study [1]. In the FDNPP case, 137Cs was transported in sulfate aerosol form [2], so the oxide aerosol behaviors could not represent the actual deposition behavior in this accident. For evaluation of whole behavior of 137Cs in vegetation system, fundamental data for deposition and uptake process of sulfate aerosol was desired. In this study, we developed aerosol generation system for simulating the dry deposition and the foliar uptake behaviors of aerosol in the different chemical constitutions. In this system, the method of aerosol generation based on the spray drying. Solution contained 137Cs was send to a nozzle by a syringe pump and spraying with a high speed air flow. The sprayed mist was generated in a chamber in the relatively high temperature. The solution in the mist was dried quickly, and micro size solid aerosols consisting 137Cs were generated. The aerosols were suctioned by an ejector and transported inside a tube by the dry air flow, then were directly blown onto the leaves. The experimental condition, such as the size of chamber, chamber temperature, solution flow rate, air flow rate and so on, were optimized. In the deposition experiment, the aerosols on leaves were observed by a SEM/EDX system and the deposition amount was evaluated by measuring the stable Cs remaining on leaf. In the presentation, we will discuss the detail

  20. Multiple Criteria Decision-Making Approach for Prioritizing Accident Hot Spots in Developing Countries in the Absence of Crash Data

    OpenAIRE

    Pirdavani, Ali; Brijs, Tom; Wets, Geert

    2010-01-01

    In an efficient transportation system, traffic safety is an important issue and it is influenced by many factors. In a country like Iran, until now safety improvements are mainly concentrated on road engineering activities, without much attention for vehicle technology or driving behavior. One important aspect of road safety engineering activities is the so-called treatment of hotspots or dangerous accident locations. Until recently, accident hotspots were identified and remedied by the exper...

  1. Spent fuel shipping cask accident evaluation

    International Nuclear Information System (INIS)

    Mathematical models have been developed to simulate the dynamic behavior, following a hypothetical accident and fire, of typical casks designed for the rail shipment of spent fuel from nuclear reactors, and to determine the extent of radioactive releases under postulated conditions. The casks modeled were the IF-300, designed by the General Electric Company for the shipment of spent LWR fuel, and a cask designed by the Aerojet Manufacturing Company for the shipment of spent LMFBR fuel

  2. Development of a multi-criteria tool to support decision-making process on decontamination of urban areas after a nuclear accident

    Energy Technology Data Exchange (ETDEWEB)

    Rochedo, Elaine R.R.; Luca, Christiano de [Instituto Militar de Engenharia, Pc. Gen. Tiburcio, 80, Praia Vermelha, Rio de Janeiro, 22290-270 RJ (Brazil); Silva, Diogo N.G. [Universidade Federal do Rio de Janeiro, Instituto de Biofisica Carlos Chagas Filho, Rio de Janeiro 21941-902 RJ (Brazil); Wasserman, Maria Angelica V. [Instituto de Engenharia Nuclear, Cidade Universitaria, Ilha do Fundao, 21941-906, Rio de Janeiro RJ (Brazil)

    2014-07-01

    This work describes the main efforts to derive criteria for classifying technical aspects related to decontamination procedures to feed a multi-criteria tool to support decisions on remediation of urban areas after nuclear accidents. After listing procedures already tested or used in previous accident, technical aspects to be considered were derived. The relevance of each aspect was determined based on questionnaires answered by experts with experience on remediation after an accident. The questionnaire included 12 aspects and for each of them one or more technical criteria where developed to allow the classification of remediation procedures for urban areas. The criteria described in this work relate to the effects of each procedure on doses to the public, doses to remediation workers, waste generation and infrastructure needed. The aim of this project was to increase public concerns by turning the decision making process more reliable and transparent. In this work, the list of criteria and associated values are described. This list is now being included in a previously developed dose assessment computer program to allow the optimization of actions to be used considering all justifiable procedures based on the current experience on dealing with urban areas contamination after a nuclear or radiological accident. (authors)

  3. Development of computer-controlled ultrasonic image processing system for severe accidents research

    Energy Technology Data Exchange (ETDEWEB)

    Koo, Kil Mo; Kang, Kyung Ho; Kim, Jong Tai; Kim, Jong Whan; Cho, Young Ro; Ha, Kwang Soon; Park, Rae Jun; Kim, Sang Baik; Kim, Hee Dong; Sim, Chul Moo

    2000-07-01

    In order to verify in-vessel corium cooling mechanism, LAVA(Lower-plenum Arrested Vessel Attack) experiment is being performed as a first stage proof of principle test. The aims of this study are to find a gap formation between corium melt and 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 developing a computer controlled image signal processing system which is able to improve visualization and to measure the gap distribution with 3-dimensional planar image using a time domain signal analysis method as a part of the ultrasonic pulse echo methods and a computerized position control system. An image signal processing system is developed by independently developing an ultrasonic image signal processing technique and a PC controlled position control system and then combining both systems.

  4. Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

    International Nuclear Information System (INIS)

    A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes

  5. Full-scale modelling of the MNSR reactor to simulate normal operation, transients and reactivity insertion accidents under natural circulation conditions using the thermal hydraulic code ATHLET

    Energy Technology Data Exchange (ETDEWEB)

    Hainoun, A. [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)]. E-mail: ahainoun@aec.org.sy; Alissa, S. [Department of Nuclear Engineering, Atomic Energy Commission, P.O. Box 6091, Damascus (Syrian Arab Republic)

    2005-01-01

    A full-scale ATHLET system model for the Syrian miniature neutron source reactor (MNSR) has been developed. The model represents all reactor components of primary and secondary loops with the corresponding neutronics and thermal hydraulic characteristics. Under the MNSR operation conditions of natural circulation, normal operation, step reactivity transients and reactivity insertion accidents have been simulated. The analyses indicate the capability of ATHLET to simulate MNSR dynamic and thermal hydraulic behaviour and particularly to calculate the core coolant velocity of prevailing natural circulation in presence of the strong negative reactivity feed back of coolant temperature. The predicted time distribution of reactor power, core inlet and outlet coolant temperature follow closely the measured data for the quasi steady and transient states. However, sensitivity analyses indicate the influence of pressure form loss coefficients at core inlet and outlet on the results. The analysis of reactivity accidents represented by the insertion of large reactivity, demonstrates the high inherent safety features of MNSR. Even in case of insertion of total available cold excess reactivity without scram, the high negative reactivity feedback of moderator temperature limits power excursion and avoids consequently the escalation of clad temperature to the level of onset of sub-cooled void formation. The calculated peak power in this case agrees well with the data reported in the safety analysis report. The ATHLET code had not previously been assessed under these conditions. The results of this comprehensive analysis ensure the ability of the code to test some conceptual design modifications of MNSR's cooling system aiming the improvement of core cooling conditions to increase the maximum continuous reactor operation time allowing more effective use of MNSR for irradiation purposes.

  6. Method for the development of emergency response preparedness for nuclear or radiological accidents

    International Nuclear Information System (INIS)

    This report supplements IAEA emergency preparedness guidance published in the 1980s, and is consistent with the new international guidance. It provides practical advice for the development of an emergency response capability based on the potential nature and magnitude of the risk. In order to apply this method, emergency planners should have a good understanding of the basic radiological emergency response principles. Therefore, other applicable international guidance should be reviewed before using this report. This report provides a practical step-by-step method for developing integrated user, local and national emergency response capabilities. It can also be used as the basis for conducting an audit of an existing emergency response capability

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

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-07-01

    The Four years of the IMPACT, 'Integrated Modular Plant Analysis and Computing Technology', project Phase 1 have been completed. At the end of Phase 1, the Basic Single-, Two-, Multi-Phase Flow Analysis Modules of Various Coordinates have been parallelized. The physical models in the Boiling Transition Analysis Code and the Fluid-Structure Interaction Analysis Code have been completed and verified by comparison with basic experimental results. The verification study of the code was conducted in two steps. First, each analysis module was run independently and analysis results were compared against separate-effect experiment data. Verification analyses included: CORA-13 (FZK) for the Core Heat-up Module; VI-3 of HI/VI Test (ORNL) for the FP Release from Fuel Module; KROTOS-37 (JRC-ISPRA) for the Molten Core Relocation Module; Water Spread Test (UCSB) for the Debris Spreading Model and Benard's Melting Test for Natural Convection Model in the Debris Cooling Module; Hydrogen Burning Test (NUPEC) for the Ex- Vessel Thermal Hydraulics Module; PREMIX, PM10 (FZK) for the Steam Explosion Module; and SWISS-2 (SNL) for the Debris-Concrete Interaction Module. All comparison showed good agreement. Second, with the Simulation Supervisory Module, these analysis modules were executed concurrently in the parallel environment to demonstrate the code capability and integrity. (J.P.N.)

  8. Culpa da vítima: um modelo para perpetuar a impunidade nos acidentes do trabalho Guilty victims: a model to perpetuate impunity for work-related accidents

    Directory of Open Access Journals (Sweden)

    Rodolfo Andrade Gouveia Vilela

    2004-04-01

    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.

  9. Containment severe accident thermohydraulic phenomena

    International Nuclear Information System (INIS)

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

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-01-01

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

  11. Development and qualification of a particulate and iodine monitor for measuring radioactive discharges in the event of accident

    International Nuclear Information System (INIS)

    In the event of accidents the German rule KTA1503.2 requires for nuclear power plants a monitoring of the radioactive effluents from the stack. In order to protect human health and the environment the amount of the radiologically relevant isotopes of iodine and particulates has to be measured over a wide range. For this object MGP Instruments has developed the PIM203 monitor with support of E.ON and other power plant companies. The monitor measures simultaneously the particulate and iodine activity. The monitor fulfills at least all requirements of the rule KTA1503.2. This was verified by a plant independent type test according to VdTUV directive 26, carried out by the German expert organization TUV Nord e. V.. Following we will present in detail the requirements of the rule KTA1503.2 and a description of the PIM203 monitor. Finally we will explain the type and extent of the qualification tests and show some relevant results. (author)

  12. Modeling the dynamics of radionuclide concentration in animal derived products after an accident in tropical areas

    International Nuclear Information System (INIS)

    Following an accidental release of radionuclides to the atmosphere with the contamination of large areas, a detailed and fast methodology to assess the prognosis of public exposure is needed to estimate radiological consequences and optimize decisions to the protection of the public. The German model ECOSYS has been chosen to integrate the SIEM - Integrated Emergency System, developed by IRD/CNEN to assess the doses to the public after an accidental contamination of rural areas. The use the model demands a considerable effort in adapting scenarios to fit the specific conditions of a location, considering the differences related to climate, environmental characteristics, agricultural calendar and practices, along with population diet. The area selected to start this adaptation considers the characteristics of the 50 km radius area surrounding the nuclear power plants at Angra dos Reis, Rio de Janeiro. At a first stage, the concentration on vegetal food products has been studied. This work describes the methodology used to select scenarios and presents results of the dynamics of the predicted concentration of radionuclides in different kinds of animal derived food products. The work provides guidance to the need of radioecological research needed to improve the adequacy of the estimates to actual Brazilian scenarios. (author)

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

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Gunhyo; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of). Dept. of Nuclear Engineering

    2015-03-15

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

  14. Economic models of compensation for damages caused by nuclear accidents: some lessons for the revision of the Paris and Vienna Conventions

    Energy Technology Data Exchange (ETDEWEB)

    Faure, Michael G. [Limburg Univ., Maastricht (Netherlands). Faculty of Law

    1995-12-31

    Alternative systems of compensation for damages caused by nuclear accidents have been proposed. In respect, the question merits attention to whether these alternative models of compensation discussed in the economic literature could be implemented when discussing the revision of the Paris and Vienna Conventions. 55 refs., 1 tab.

  15. Comments on the A. Aurengo document entitled 'the 2003 I.R.S.N. model of Chernobylsk accident fallout in France' - 18 december 2005

    International Nuclear Information System (INIS)

    The present document brings the principal comments of I.R.S.N. on the analysis note of the 18. december 2005 treating the 2003 I.R.S.N. model on the Chernobylsk accident fallout in France, transmitted by Mister A. Aurengo. (N.C.)

  16. Development of two mobile laboratories for a routine and accident monitoring of internal contamination

    Energy Technology Data Exchange (ETDEWEB)

    Franck, D., E-mail: didier.franck@irsn.fr [IRSN-French Institute for Radiological Protection and Nuclear Safety, Internal Dosimetry Department, BP 17, 92262 Fontenay-aux-Roses Cedex (France); Berniere, J.; Viltard, D.; Parre, F.; Challeton-de Vathaire, C.; Agarande, M. [IRSN-French Institute for Radiological Protection and Nuclear Safety, Internal Dosimetry Department, BP 17, 92262 Fontenay-aux-Roses Cedex (France)

    2012-07-15

    To provide medical surveillance of workers exposed to risk of internal contamination, IRSN (French Institute for Radiological Protection and Nuclear Safety) has developed two mobile laboratories for on-site monitoring. The laboratories are unique in Europe. They meet the new radiation protection requirements for nuclear medicine departments and radiological emergency response. Details of the design, calibration procedures and performance characteristics of these systems in measurements of various types of organs (thyroid, lung and whole body) are described. The sensitivity of the measurements is very close to that achieved in a heavily shielded stationary laboratory. - Highlights: Black-Right-Pointing-Pointer We describe a new mobile truck developed for in situ monitoring of internal contamination. Black-Right-Pointing-Pointer We present the technical principle of this truck. Black-Right-Pointing-Pointer They can carry out in vivo measurements in different types of geometries: thyroid, lung and whole body. Black-Right-Pointing-Pointer We assess the performances for the measurement of typical radionuclides and contaminations. Black-Right-Pointing-Pointer Thanks to the use of specific shielding it has been shown sensitivities quite close to a heavy shielding system.

  17. Influence of the meteorological input on the atmospheric transport modelling with FLEXPART of radionuclides from the Fukushima Daiichi nuclear accident

    International Nuclear Information System (INIS)

    In the present paper the role of precipitation as FLEXPART model input is investigated for one possible release scenario of the Fukushima Daiichi accident. Precipitation data from the European Center for Medium-Range Weather Forecast (ECMWF), the NOAA's National Center for Environmental Prediction (NCEP), the Japan Meteorological Agency's (JMA) mesoscale analysis and a JMA radar-rain gauge precipitation analysis product were utilized. The accident of Fukushima in March 2011 and the following observations enable us to assess the impact of these precipitation products at least for this single case. As expected the differences in the statistical scores are visible but not large. Increasing the ECMWF resolution of all the fields from 0.5° to 0.2° rises the correlation from 0.71 to 0.80 and an overall rank from 3.38 to 3.44. Substituting ECMWF precipitation, while the rest of the variables remains unmodified, by the JMA mesoscale precipitation analysis and the JMA radar gauge precipitation data yield the best results on a regional scale, specially when a new and more robust wet deposition scheme is introduced. The best results are obtained with a combination of ECMWF 0.2° data with precipitation from JMA mesoscale analyses and the modified wet deposition with a correlation of 0.83 and an overall rank of 3.58. NCEP-based results with the same source term are generally poorer, giving correlations around 0.66, and comparatively large negative biases and an overall rank of 3.05 that worsens when regional precipitation data is introduced

  18. Accident-precipitating factors for crashes in turbine-powered general aviation aircraft.

    Science.gov (United States)

    Boyd, Douglas D; Stolzer, Alan

    2016-01-01

    General aviation (14CFR Part 91) accounts for 83% of civil aviation fatalities. While much research has focused on accident causes/pilot demographics in this aviation sector, studies to identify factors leading up to the crash (accident-precipitating factors) are few. Such information could inform on pre-emptive remedial action. With this in mind and considering the paucity of research on turbine-powered aircraft accidents the study objectives were to identify accident-precipitating factors and determine if the accident rate has changed over time for such aircraft operating under 14CFR Part 91. The NTSB Access database was queried for accidents in airplanes (engines and occurring between 1989 and 2013. We developed and utilized an accident-precipitating factor taxonomy. Statistical analyses employed logistic regression, contingency tables and a generalized linear model with Poisson distribution. The "Checklist/Flight Manual Not Followed" was the most frequent accident-precipitating factor category and carried an excess risk (OR 2.34) for an accident with a fatal and/or serious occupant injury. This elevated risk reflected an over-representation of accidents with fatal and/or serious injury outcomes (pengine aircraft are more frequent than their single engine counterparts and the decline (50%) in the turbine aircraft accident rate over the study period was likely due, in part, to a 6-fold increased representation of single engine airplanes. In conclusion, our study is the first to identify novel precursive factors for accidents involving turbine aircraft operating under 14CFR Part 91. This research highlights areas that should receive further emphasis in training/recurrency in a pre-emptive attempt to nullify candidate accident-precipitating factor(s).

  19. Accident-precipitating factors for crashes in turbine-powered general aviation aircraft.

    Science.gov (United States)

    Boyd, Douglas D; Stolzer, Alan

    2016-01-01

    General aviation (14CFR Part 91) accounts for 83% of civil aviation fatalities. While much research has focused on accident causes/pilot demographics in this aviation sector, studies to identify factors leading up to the crash (accident-precipitating factors) are few. Such information could inform on pre-emptive remedial action. With this in mind and considering the paucity of research on turbine-powered aircraft accidents the study objectives were to identify accident-precipitating factors and determine if the accident rate has changed over time for such aircraft operating under 14CFR Part 91. The NTSB Access database was queried for accidents in airplanes (turbine engines and occurring between 1989 and 2013. We developed and utilized an accident-precipitating factor taxonomy. Statistical analyses employed logistic regression, contingency tables and a generalized linear model with Poisson distribution. The "Checklist/Flight Manual Not Followed" was the most frequent accident-precipitating factor category and carried an excess risk (OR 2.34) for an accident with a fatal and/or serious occupant injury. This elevated risk reflected an over-representation of accidents with fatal and/or serious injury outcomes (pturbine aircraft accident rate over the study period was likely due, in part, to a 6-fold increased representation of single engine airplanes. In conclusion, our study is the first to identify novel precursive factors for accidents involving turbine aircraft operating under 14CFR Part 91. This research highlights areas that should receive further emphasis in training/recurrency in a pre-emptive attempt to nullify candidate accident-precipitating factor(s). PMID:26590507

  20. Comparison of Conceptual Graphs for Modelling Knowledge of Multiple Experts : Application to Traffic Accident Analysis

    OpenAIRE

    Dieng, Rose

    1997-01-01

    When modelling knowledge of multiple experts, it is interesting to build a common expertise model corresponding to the kernel of knowledge common to the experts. Therefore the expertise conflicts between the expertise models of the different experts must be tackled. The domain level of an expertise model can be described through concepts linked by relations, and represented through Sowa's conceptual graph formalism. This report presents a method for conflict management during knowledge modell...

  1. Internal Flow Thermal/Fluid Modeling of STS-107 Port Wing in Support of the Columbia Accident Investigation Board

    Science.gov (United States)

    Sharp, John R.; Kittredge, Ken; Schunk, Richard G.

    2003-01-01

    As part of the aero-thermodynamics team supporting the Columbia Accident Investigation Board (CAB), the Marshall Space Flight Center was asked to perform engineering analyses of internal flows in the port wing. The aero-thermodynamics team was split into internal flow and external flow teams with the support being divided between shorter timeframe engineering methods and more complex computational fluid dynamics. In order to gain a rough order of magnitude type of knowledge of the internal flow in the port wing for various breach locations and sizes (as theorized by the CAB to have caused the Columbia re-entry failure), a bulk venting model was required to input boundary flow rates and pressures to the computational fluid dynamics (CFD) analyses. This paper summarizes the modeling that was done by MSFC in Thermal Desktop. A venting model of the entire Orbiter was constructed in FloCAD based on Rockwell International s flight substantiation analyses and the STS-107 reentry trajectory. Chemical equilibrium air thermodynamic properties were generated for SINDA/FLUINT s fluid property routines from a code provided by Langley Research Center. In parallel, a simplified thermal mathematical model of the port wing, including the Thermal Protection System (TPS), was based on more detailed Shuttle re-entry modeling previously done by the Dryden Flight Research Center. Once the venting model was coupled with the thermal model of the wing structure with chemical equilibrium air properties, various breach scenarios were assessed in support of the aero-thermodynamics team. The construction of the coupled model and results are presented herein.

  2. Statistical modelling of traffic safety development

    DEFF Research Database (Denmark)

    Christens, Peter

    2004-01-01

    Road safety is a major concern for society and individuals. Although road safety has improved in recent years, the number of road fatalities is still unacceptably high. In 2000, road accidents killed over 40,000 people in the European Union and injured more than 1.7 million. In 2001 in Denmark...... there were 6861 injury trafficc accidents reported by the police, resulting in 4519 minor injuries, 3946 serious injuries, and 431 fatalities. The general purpose of the research was to improve the insight into aggregated road safety methodology in Denmark. The aim was to analyse advanced statistical methods......, that were designed to study developments over time, including effects of interventions. This aim has been achieved by investigating variations in aggregated Danish traffic accident series and by applying state of the art methodologies to specific case studies. The thesis comprises an introduction...

  3. Development of a phenomenological constitutive model for polyurethane foams

    International Nuclear Information System (INIS)

    Rigid, closed-cell, polyurethane foam is used in impact limiters in nuclear waste transport containers. During a hypothetical nuclear waste transport accident, the foam is expected to absorb a significant amount of impact energy by undergoing large inelastic volume reductions. Consequently, the crushing of polyurethane foams must be well characterized and accurately modeled to properly analyze a transport container accident. At the request of Sandia National Laboratories, a series of uniaxial, hydrostatic and triaxial compression tests on polyurethane foams were performed by the New Mexico Engineering Research Institute (NMERI). The combination of hydrostatic and triaxial tests was chosen to provide sufficient data to characterize both the volumetric and deviatoric behaviors of the foams and the coupling between the two responses. Typical results from the NMERI tests are included in this paper. A complete description of these tests can be found in Neilsen et al., 1987. Constitutive models that have been used in the past to model foam did not capture some important foam behaviors observed in the NMERI tests. Therefore, a new constitutive model for rigid, closed-cell, polyurethane foams was developed and implemented in two finite element codes. Development of the new model is discussed in this paper. Also, results from analyses with the new model and other constitutive models are presented to demonstrate differences between the various models. 4 refs., 6 figs., 1 tab

  4. Development of a phenomenological constitutive model for polyurethane foams

    International Nuclear Information System (INIS)

    Rigid, closed-cell, polyurethane foam is used in impact limiters in nuclear waste transport containers. During a hypothetical nuclear waste transport accident, the foam is expected to absorb a significant amount of impact energy by undergoing large inelastic volume reductions. Consequently, the crushing of polyurethane foams must be well characterized and accurately modeled to properly analyze a transport container accident. At the request of Sandia National Laboratories, a series of uniaxial, hydrostatic and triaxial compression tests on polyurethane foams were performed by the New Mexico Engineering Research Institute (NMERI). The combination of hydrostatic and triaxial tests was chosen to provide sufficient data to characterize both the volumetric and deviatoric behaviors of the foams and the coupling between the two responses. Typical results from the NMERI tests are included in this paper. Constitutive models that have been used in the past to model foam did not capture some important foam behaviors observed in the NMERI tests. Therefore, a new constitutive model for rigid, closed-cell, polyurethane foams was developed and implemented in two finite element codes. Development of the new model is discussed in this paper. Also, results from analyses with the new model and other constitutive models are presented to demonstrate differences between the various models

  5. An international standard problem: analysis of 1:4-scale prestressed concrete containment vessel model under severe accident conditions

    International Nuclear Information System (INIS)

    In June 2002, The OECD-NEA Committee on the Safety of Nuclear Installations (CSNI), with the encouragement of the US NRC, initiated an International Standard Problem on containment integrity (ISP 48) based on the NRC/NUPEC/Sandia test. The objectives of the ISP are to extend the understanding of capacities of actual containment structures based on results of the recent PCCV Model test and other previous research. From 1997 through 2001 Sandia National Laboratories (SNL) conducted a Cooperative Containment Integrity Program under the joint sponsorship of the Nuclear Power Engineering Corporation (NUPEC) of Japan, and the NRC Office of Nuclear Regulatory Research. The purpose of the program was to investigate the response of representative models of nuclear containment structures to pressure loading beyond the design basis accident and to compare analytical predictions to measured behavior. A uniform 1:4-scale model of a prestressed concrete containment vessel (PCCV) was constructed and tested at SNL. This model was representative of the containment structure of an actual pressurized-water reactor plant in Japan. The ISP consists of four phases over a period of 2 years: Phase 1: Data Collection and Identification Phase 2: Calculation of the Limit State Test (LST), i.e. static pressure loading Phase 3: Calculation of response to both Thermal and Mechanical Loadings Phase 4: Reporting Workshop Eleven organizations (or teams) from nine OECD member countries accepted the invitation to participate in the ISP and perform calculations to predict the structural response of the PCCV model to static and transient pressure and thermal loading. Each participating organization was provided with the model and loading data and was asked to perform independent analyses to simulate the response of the PCCV model. The results of each team's calculations were compiled and the results presented at a final workshop in April 2005. These results and the conclusions and insights gained from

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

    International Nuclear Information System (INIS)

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

  7. Accident Prediction Models for Urban Arterial System%城市干道系统交通事故预测模型研究

    Institute of Scientific and Technical Information of China (English)

    孟祥海; 陈天恩; 盛洪飞; 姜美利

    2007-01-01

    It relies greatly on the accident prediction models to make effective traffic safety countermeasures. Therefore, by taking Harbin urban arterial network composed of 468 arterial links and 163 at-grade intersections as a case, the broad geometry and traffic flow data of the network were collected, as well as 8 510 accident data occurred on the network during 1999 to 2004. Firstly, the characterist ics of the traffic accident data were analyzed, and the results show that theaccident data follow the Negative Binomial distribution. Secondly, links and inter sections were classified according to the cluster analysis method, and then the accident prediction models that can be used to predict the accident frequencies occurred on each kind of links and intersections were established. Thirdly, the quantitative relationship between the accident index of the links during rushh ours and the v/c of them was discussed. Totally, 24 prediction models were calibrated. Finally, the prediction models were applied to a case study on partial road network of Harbin, which was planned for the target year of 2010. There sults show the fact that the accident prediction models are effective.%以哈尔滨市干道路网为研究对象,收集到了该路网上468个路段和163个平面交叉口的道路交通数据,以及1999年至2004年所发生的8510起交通事故数据.分析了事故数据的统计分布特性,应用聚类分析技术确定了路段和交叉口的类别 ,并在此基础上分别建立了事故总体和分事故形态的预测模型.论文探讨了高峰时段的事故次数、事故率与路段v/c之间的定量关系.标定出了24个模型,并形成干道系统事故预测模型库.最后,运用所建立的事故预测模型选取了2010年哈尔滨规划路网的一部分进行实例分析,结果表明了预测模型是有效的.

  8. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of atmospheric dispersion model with improved deposition scheme and oceanic dispersion model

    OpenAIRE

    G. Katata; Chino, M; T. Kobayashi; Terada, H.; Ota, M; Nagai, H.; M. Kajino; Draxler, R; M. C. Hort; Malo, A.; Torii, T.; Y. Sanada

    2014-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Dai-ichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate a detailed time trend of atmospheric releases during the accident by combining environmental monitoring data with atmospheric model simulations from WSP...

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

    International Nuclear Information System (INIS)

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

  10. Models and criteria for prediction of Deflagration-to-Detonation Transition (DDT) in hydrogen-air-steam systems under severe accident conditions. Final report

    International Nuclear Information System (INIS)

    The European Commission in Brussels supported a joint project on Deflagration-to-Detonation Transition (DDT) studies for hydrogen safety within the framework programme on nuclear fission safety. The project was initiated by the Forschungszentrum Juelich based on the results of a pilot project. The following main project was coordinated by the Freie Universitaet Berlin involving seven european partners. The partners came from universities, research centers and industry, as follows: FU-Berlin, RWTH-Aachen, CNRS-Marseille, IPSN-Saclay, FZ-Juelich, FZ-Karlsruhe, and NNC-Knutsford, which worked closely together. The working period was two years (1997-1998). The aim of the project was to develop models and criteria for prediction of deflagration-to-detonation transition (DDT) in hydrogen-air-steam systems under severe accident conditions. The results obtained are documented in this final report, which was finished in 1999. The report consists of seven chapters, concerning: - Introduction - Experimental Investigations - Modelling and Numerics - Validation - Mitigation - Further Deliverables - Summary and Conclusion. The final report presents special experimental, theoretical, and computational aspects of the complex DDT phenomena for hydrogen safety studies, and it should be a solid basis for end user applications and further developments. (orig.)

  11. ATMES: An evaluation of long range atmospheric dispersion models using radioactivity data from the Chernobyl accident

    International Nuclear Information System (INIS)

    ATMES was organized in 1989 by CEC, WMO and IAEA with the main purpose of reviewing and calibrating long range atmospheric dispersion models, to identify the more successful approaches. This document provides some findings of the study which employed 21 models from various institutions of 14 countries. They had to produce the model results without knowing in advance the measurements to which their results had to be compared. Many out of the participating models were able to predict the arrival times at the two selected sites using the analysed windfields within a few hours

  12. Analysis and model testing of Super Tiger Type B packaging in accident environments

    International Nuclear Information System (INIS)

    Based on previous scale model test results with more rigid systems and the subsystem tests on drums, it is believed that the scaled models realistically replicate full scale system behavior. Future work will be performed to obtain improved stiffness data on the Type A containers. These data will be incorporated into the finite element model, and improved correlation with the test results is expected. Review of the scale model transport system test results indicated that the method of attachment of the Super Tiger to the trailer was the primary cause for detachment of the outer door during the one-eighth scale grade-crossing test. Although the container seal on the scale model of Super Tiger was not adequately modeled to provide a leak-tight seal, loss of the existing seal in a full scale test can be inferred from the results of the one-quarter scale model grade-crossing test. In each test, approximately two-thirds of the model drums were estimated to have deformed sufficiently to predict loss of drum head closure seal, with several partially losing their contents within the overpack. In no case were drums ejected from the overpack, nor was there evidence of material loss in excess of the amount assumed in the WIPP EIS from any of the Super Tiger models tested. 9 figures

  13. Scoping accident(s) for emergency planning

    International Nuclear Information System (INIS)

    At the request of the Conference of State Radiation Control Program Director's (CRCPD), in November 1976 the U.S. Nuclear Regulatory Commission formed a joint Task Force with representatives of the U.S. Environmental Protection Agency to answer a number of questions posed by the States regarding emergency planning. This Task Force held monthly meetings through November 1977. In December 1977 a draft report was prepared for limited distribution for review and comment by selected State and local organizations. The NRC/EPA Task Force deliberations centered on the CRCPD request for '... a determination of the most severe accident basis for which radiological emergency response plans should be developed by offsite agencies...' in the vicinity of nuclear power plants. Federal Interagency guidance to the States in this regard has been that the scoping accident should be the most serious conservatively analyzed accident considered for siting purposes, as exemplified in the Commission's Regulations at 10 CFR Part 100 and the NRC staffs Regulatory Guides 1.3 and 1.4, and as presented in license applicant's Safety Analysis Reports and the USNRC Staffs Safety Evaluation Reports. The draft report of the Task Force amplifies on this recommendation: to present a clearer picture of its import and introduces the concept of protective action zones (PAZs) within which detailed emergency plans should be developed; one zone for the plume exposure pathway and a second, larger zone for contamination pathways. The time dependence of potential releases and atmospheric transport, and important radionuclide groups of possible import are also discussed in the draft Task Force report. A status report regarding this effort, as of June 1978, will be presented. (author)

  14. Desktop Severe Accident Graphic Simulator Module for CANDU6 : PSAIS

    Energy Technology Data Exchange (ETDEWEB)

    Park, S. Y.; Song, Y. M. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2015-10-15

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

  15. Estimating the Influence of Accident Related Factors on Motorcycle Fatal Accidents using Logistic Regression (Case Study: Denpasar-Bali

    Directory of Open Access Journals (Sweden)

    Wedagama D.M.P.

    2010-01-01

    Full Text Available In Denpasar the capital of Bali Province, motorcycle accident contributes to about 80% of total road accidents. Out of those motorcycle accidents, 32% are fatal accidents. This study investigates the influence of accident related factors on motorcycle fatal accidents in the city of Denpasar during period 2006-2008 using a logistic regression model. The study found that the fatality of collision with pedestrians and right angle accidents were respectively about 0.44 and 0.40 times lower than collision with other vehicles and accidents due to other factors. In contrast, the odds that a motorcycle accident will be fatal due to collision with heavy and light vehicles were 1.67 times more likely than with other motorcycles. Collision with pedestrians, right angle accidents, and heavy and light vehicles were respectively accounted for 31%, 29%, and 63% of motorcycle fatal accidents.

  16. Risk analysis of emergent water pollution accidents based on a Bayesian Network.

    Science.gov (United States)

    Tang, Caihong; Yi, Yujun; Yang, Zhifeng; Sun, Jie

    2016-01-01

    To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents.

  17. Risk analysis of emergent water pollution accidents based on a Bayesian Network.

    Science.gov (United States)

    Tang, Caihong; Yi, Yujun; Yang, Zhifeng; Sun, Jie

    2016-01-01

    To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents. PMID:26433361

  18. SOURCE 2.0 model development: UO{sub 2} thermal properties

    Energy Technology Data Exchange (ETDEWEB)

    Reid, P.J. [ALARA Research, Inc., Saint John, New Brunswick (Canada); Richards, M.J. [Hydro Quebec, Montreal, Quebec (Canada); Iglesias, F.C.; Brito, A.C. [Ontario Hydro, Toronto, Ontario (Canada)

    1997-07-01

    During analysis of CANDU postulated accidents, the reactor fuel is estimated to experience large temperature variations and to be exposed to a variety of environments from highly oxidized to mildly reducing. The exposure of CANDU fuel to these environments and temperatures may affect fission product releases from the fuel and cause degradation of the fuel thermal properties. The SOURCE 2.0 project is a safety analysis code which will model the necessary mechanisms required to calculate fission product release for a variety of accident scenarios, including large break loss of coolant accidents (LOCAs) with or without emergency core cooling. The goal of the model development is to generate models which are consistent with each other and phenomenologically based, insofar as that is possible given the state of theoretical understanding.

  19. Severe Accident Recriticality Analyses (SARA)

    Energy Technology Data Exchange (ETDEWEB)

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

    1999-11-01

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

  20. Light-Weight Radioisotope Heater Unit Safety Analysis Report (LWRHU-SAR). Volume II. Accident model document

    International Nuclear Information System (INIS)

    Purposes of this volume (AMD), are to: Identify all malfunctions, both singular and multiple, which can occur during the complete mission profile that could lead to release outside the clad of the radioisotopic material contained therein; provide estimates of occurrence probabilities associated with these various accidents; evaluate the response of the LWRHU (or its components) to the resultant accident environments; and associate the potential event history with test data or analysis to determine the potential interaction of the released radionuclides with the biosphere

  1. Water reactor fuel element computer modelling in steady state, transient and accident conditions

    International Nuclear Information System (INIS)

    The present meeting was scheduled by the International Atomic Energy Agency, upon proposal of the Members of the International Working Group on Water Reactor Fuel Performance and Technology (IWGFPT). This meeting was the fifth in the series of IAEA meetings on the topic of Water Reactor Fuel Element Modelling, previous meetings being held in 1978, 1980, 1982 and 1984. Sixty-seven participants from 21 countries attended the meeting, and 35 papers were presented and discussed. These numbers are almost exactly the same as for the 1984 meeting, which demonstrates a continuing interest in the topic. The papers were presented in five sessions under the following headings: Session I - General Modelling (6 papers); Session II - Thermo-Mechanical Modelling and PCI (7 papers); Session III - Fission Gas Release (7 papers); Session IV - Transient Behaviour (8 papers); Session V - Axial Gas Transport and Thermal Modelling (7 papers). A separate abstract was prepared for each of these 35 papers. Refs, figs and tabs

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

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

    International Nuclear Information System (INIS)

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

  5. Emulation and Sobol' sensitivity analysis of an atmospheric dispersion model applied to the Fukushima nuclear accident

    Science.gov (United States)

    Girard, Sylvain; Mallet, Vivien; Korsakissok, Irène; Mathieu, Anne

    2016-04-01

    Simulations of the atmospheric dispersion of radionuclides involve large uncertainties originating from the limited knowledge of meteorological input data, composition, amount and timing of emissions, and some model parameters. The estimation of these uncertainties is an essential complement to modeling for decision making in case of an accidental release. We have studied the relative influence of a set of uncertain inputs on several outputs from the Eulerian model Polyphemus/Polair3D on the Fukushima case. We chose to use the variance-based sensitivity analysis method of Sobol'. This method requires a large number of model evaluations which was not achievable directly due to the high computational cost of Polyphemus/Polair3D. To circumvent this issue, we built a mathematical approximation of the model using Gaussian process emulation. We observed that aggregated outputs are mainly driven by the amount of emitted radionuclides, while local outputs are mostly sensitive to wind perturbations. The release height is notably influential, but only in the vicinity of the source. Finally, averaging either spatially or temporally tends to cancel out interactions between uncertain inputs.

  6. Detailed source term estimation of the atmospheric release for the Fukushima Daiichi Nuclear Power Station accident by coupling simulations of an atmospheric dispersion model with an improved deposition scheme and oceanic dispersion model

    Science.gov (United States)

    Katata, G.; Chino, M.; Kobayashi, T.; Terada, H.; Ota, M.; Nagai, H.; Kajino, M.; Draxler, R.; Hort, M. C.; Malo, A.; Torii, T.; Sanada, Y.

    2015-01-01

    Temporal variations in the amount of radionuclides released into the atmosphere during the Fukushima Daiichi Nuclear Power Station (FNPS1) accident and their atmospheric and marine dispersion are essential to evaluate the environmental impacts and resultant radiological doses to the public. In this paper, we estimate the detailed atmospheric releases during the accident using a reverse estimation method which calculates the release rates of radionuclides by comparing measurements of air concentration of a radionuclide or its dose rate in the environment with the ones calculated by atmospheric and oceanic transport, dispersion and deposition models. The atmospheric and oceanic models used are WSPEEDI-II (Worldwide version of System for Prediction of Environmental Emergency Dose Information) and SEA-GEARN-FDM (Finite difference oceanic dispersion model), both developed by the authors. A sophisticated deposition scheme, which deals with dry and fog-water depositions, cloud condensation nuclei (CCN) activation, and subsequent wet scavenging due to mixed-phase cloud microphysics (in-cloud scavenging) for radioactive iodine gas (I2 and CH3I) and other particles (CsI, Cs, and Te), was incorporated into WSPEEDI-II to improve the surface deposition calculations. The results revealed that the major releases of radionuclides due to the FNPS1 accident occurred in the following periods during March 2011: the afternoon of 12 March due to the wet venting and hydrogen explosion at Unit 1, midnight of 14 March when the SRV (safety relief valve) was opened three times at Unit 2, the morning and night of 15 March, and the morning of 16 March. According to the simulation results, the highest radioactive contamination areas around FNPS1 were created from 15 to 16 March by complicated interactions among rainfall, plume movements, and the temporal variation of release rates. The simulation by WSPEEDI-II using the new source term reproduced the local and regional patterns of cumulative

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

    Energy Technology Data Exchange (ETDEWEB)

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

    2015-08-01

    Accident-tolerant fuels (ATFs) are fuels and/or cladding that, in comparison with the standard uranium dioxide Zircaloy system, can tolerate loss of active cooling in the core for a considerably longer time period while maintaining or improving the fuel performance during normal operations [1]. It is important to note that the currently used uranium dioxide Zircaloy fuel system tolerates design basis accidents (and anticipated operational occurrences and normal operation) as prescribed by the US Nuclear Regulatory Commission. Previously, preliminary simulations of the plant response have been performed under a range of accident scenarios using various ATF cladding concepts and fully ceramic microencapsulated fuel. Design basis loss of coolant accidents (LOCAs) and station blackout (SBO) severe accidents were analyzed at Oak Ridge National Laboratory (ORNL) for boiling water reactors (BWRs) [2]. Researchers have investigated the effects of thermal conductivity on design basis accidents [3], investigated silicon carbide (SiC) cladding [4], as well as the effects of ATF concepts on the late stage accident progression [5]. These preliminary analyses were performed to provide initial insight into the possible improvements that ATF concepts could provide and to identify issues with respect to modeling ATF concepts. More recently, preliminary analyses for a range of ATF concepts have been evaluated internationally for LOCA and severe accident scenarios for the Chinese CPR1000 [6] and the South Korean OPR-1000 [7] pressurized water reactors (PWRs). In addition to these scoping studies, a common methodology and set of performance metrics were developed to compare and support prioritizing ATF concepts [8]. A proposed ATF concept is based on iron-chromium-aluminum alloys (FeCrAl) [9]. With respect to enhancing accident tolerance, FeCrAl alloys have substantially slower oxidation kinetics compared to the zirconium alloys typically employed. During a severe accident, Fe

  8. The I.R.S.N.-2003 model of Chernobylsk accident fallout in France

    International Nuclear Information System (INIS)

    In conclusion, the whole of the method impreciseness, the sometime important variation between the very simplifying hypothesis used and the available data ( especially for the air contamination and the fallout of nuclear tests), the variations between the results of the model and the measurements, its incompatible results with these ones exposed and validated in the report I.P.S.N.-1997, lead to consider that the I.R.S.N.-2003 model does not allow a reliable quantitative estimation of Chernobylsk fallout and let alone the dosimetry estimations for which it has been used. (N.C.)

  9. Exploratory analysis of Spanish energetic mining accidents.

    Science.gov (United States)

    Sanmiquel, Lluís; Freijo, Modesto; Rossell, Josep M

    2012-01-01

    Using data on work accidents and annual mining statistics, the paper studies work-related accidents in the Spanish energetic mining sector in 1999-2008. The following 3 parameters are considered: age, experience and size of the mine (in number of workers) where the accident took place. The main objective of this paper is to show the relationship between different accident indicators: risk index (as an expression of the incidence), average duration index for the age and size of the mine variables (as a measure of the seriousness of an accident), and the gravity index for the various sizes of mines (which measures the seriousness of an accident, too). The conclusions of this study could be useful to develop suitable prevention policies that would contribute towards a decrease in work-related accidents in the Spanish energetic mining industry. PMID:22721539

  10. Dynamics models and modeling of tree stand development

    Directory of Open Access Journals (Sweden)

    M. V. Rogozin

    2015-04-01

    Full Text Available Brief analysis of scientific works in Russia and in the CIS over the past 100 years. Logical and mathematical models consider the conceptual and show some of the results of their verification. It was found that the models include different laws and the parameters, the sum of which allows you to divide them into four categories: models of static states, development models, models of care for the natural forest and models of cultivation. Each category has fulfilled and fulfills its tasks in economic management. Thus, the model states in statics (table traverse growth played a prominent role in figuring out what may be the most productive (full stands in different regions of the country. However, they do not answer the question of what the initial states lead to the production of complete stands. In a study of the growth of stands used system analysis, and it is observed dominance of works studying static state, snatched from the biological time. Therefore, the real drama of the growth of stands remained almost unexplored. It is no accident there were «chrono-forestry» «plantation forestry» and even «non-traditional forestry», where there is a strong case of a number of new concepts of development stands. That is quite in keeping with Kuhn (Kuhn, 2009 in the forestry crisis began – there were alternative theories and coexist conflicting scientific schools. To develop models of stand development, it is proposed to use a well-known method of repeated observations within 10–20 years, in conjunction with the explanation of the history of the initial density. It mounted on the basis of studying the dynamics of its indicators: the trunk, crown overlap coefficient, the sum of volumes of all crowns and the relative length of the crown. According to these indicators, the researcher selects natural series of development stands with the same initial density. As a theoretical basis for the models it is possible to postulate the general properties of

  11. Modelling for great breaks accident analysis in the primary system of Angra 1 reactor

    International Nuclear Information System (INIS)

    An analysis is made for a break in the cold leg, of the guillotine type with discharge coefficient C sub(D)=1.0, for the Angra 1 reactor. The computer codes, geometrical models and options used are described. A comparison between the method used and the requirements in the Appendix K of 10 CRF 50 is done. (Author)

  12. World Meteorological Organization's model simulations of the radionuclide dispersion and deposition from the Fukushima Daiichi nuclear power plant accident.

    Science.gov (United States)

    Draxler, Roland; Arnold, Dèlia; Chino, Masamichi; Galmarini, Stefano; Hort, Matthew; Jones, Andrew; Leadbetter, Susan; Malo, Alain; Maurer, Christian; Rolph, Glenn; Saito, Kazuo; Servranckx, René; Shimbori, Toshiki; Solazzo, Efisio; Wotawa, Gerhard

    2015-01-01

    Five different atmospheric transport and dispersion model's (ATDM) deposition and air concentration results for atmospheric releases from the Fukushima Daiichi nuclear power plant accident were evaluated over Japan using regional (137)Cs deposition measurements and (137)Cs and (131)I air concentration time series at one location about 110 km from the plant. Some of the ATDMs used the same and others different meteorological data consistent with their normal operating practices. There were four global meteorological analyses data sets available and two regional high-resolution analyses. Not all of the ATDMs were able to use all of the meteorological data combinations. The ATDMs were configured identically as much as possible with respect to the release duration, release height, concentration grid size, and averaging time. However, each ATDM retained its unique treatment of the vertical velocity field and the wet and dry deposition, one of the largest uncertainties in these calculations. There were 18 ATDM-meteorology combinations available for evaluation. The deposition results showed that even when using the same meteorological analysis, each ATDM can produce quite different deposition patterns. The better calculations in terms of both deposition and air concentration were associated with the smoother ATDM deposition patterns. The best model with respect to the deposition was not always the best model with respect to air concentrations. The use of high-resolution mesoscale analyses improved ATDM performance; however, high-resolution precipitation analyses did not improve ATDM predictions. Although some ATDMs could be identified as better performers for either deposition or air concentration calculations, overall, the ensemble mean of a subset of better performing members provided more consistent results for both types of calculations.

  13. A general approach to critical infrastructure accident consequences analysis

    Science.gov (United States)

    Bogalecka, Magda; Kołowrocki, Krzysztof; Soszyńska-Budny, Joanna

    2016-06-01

    The probabilistic general model of critical infrastructure accident consequences including the process of the models of initiating events generated by its accident, the process of environment threats and the process of environment degradation is presented.

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

    International Nuclear Information System (INIS)

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

  15. Models for estimation of the radiation exposure in the case of incidents and accidents

    International Nuclear Information System (INIS)

    Appropriate radioecological and dosimetric models for the paths of distribution of the released activities in the atmosphere and in ground-level air serve the estimation of the inhalation and submersion dose rate as well as of the ingestition dose rate via food and drinking water. A release of activity off a final storage site in a salt dome does not involve a direct contamination of the biosphere. (DG)

  16. Reflection on a model of accident reporting to help to implement efficient prevention strategies

    OpenAIRE

    Textoris, Raoul; Tanzi, Tullio Joseph

    2013-01-01

    International audience In our modern societies, socio-technological systems and human system interactions are taking on a large part in numerous domains such as health, control of risk, people safety, communication, information technologies, and so on. In order to manage such systems, it is necessary to put in place the most relevant actions and indicators. To facilitate decision making in various fields, such as people safety and risk management, the use of appropriate model and the defin...

  17. Development and operation of the network system for nuclear safety - Improvement of the following accident dose assessment system

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Enn Han; Han, Moon Hee; Suh, Kyung Suk; Hwang, Won Tae [Korea Atomic Energy Research Institute, Taejon (Korea)

    1999-12-01

    The FADAS has been updates for calculating the real-time wind fields continuously at the nuclear sites in Korea. The system has been constructed to compute the wind fields using its own process for the dummy meteorological data, and does not effect on the overall wind field module. If the radioactive materials are released into the atmosphere in real situation, the calculations of wind fields and exposure dose in the previous FADAS are performed in the case of the recognition of the above situation in the source term evaluation module. The current version of FADAS includes the program for evaluating the effect of the predicted accident and the assumed scenario together. The dose assessment module is separated into the real-time and the supposed accident respectively. 8 refs., 17 figs., 10 tabs. (Author)

  18. A real-time, dynamic early-warning model based on uncertainty analysis and risk assessment for sudden water pollution accidents.

    Science.gov (United States)

    Hou, Dibo; Ge, Xiaofan; Huang, Pingjie; Zhang, Guangxin; Loáiciga, Hugo

    2014-01-01

    A real-time, dynamic, early-warning model (EP-risk model) is proposed to cope with sudden water quality pollution accidents affecting downstream areas with raw-water intakes (denoted as EPs). The EP-risk model outputs the risk level of water pollution at the EP by calculating the likelihood of pollution and evaluating the impact of pollution. A generalized form of the EP-risk model for river pollution accidents based on Monte Carlo simulation, the analytic hierarchy process (AHP) method, and the risk matrix method is proposed. The likelihood of water pollution at the EP is calculated by the Monte Carlo method, which is used for uncertainty analysis of pollutants' transport in rivers. The impact of water pollution at the EP is evaluated by expert knowledge and the results of Monte Carlo simulation based on the analytic hierarchy process. The final risk level of water pollution at the EP is determined by the risk matrix method. A case study of the proposed method is illustrated with a phenol spill accident in China. PMID:24781332

  19. Wireless. Automation of accident mountings. Development of a new solution of drive; Kabellos. Automatisierung von Havariearmaturen. Entwicklung einer neuen Antriebsloesung

    Energy Technology Data Exchange (ETDEWEB)

    Sacharowitz, Axel [3S Antriebe GmbH, Berlin (Germany)

    2013-01-15

    The Industrielle Werke Basel IWB (Basel, Switzerland) evaluate the automation of dividing an accident of its district heating grid. However, only a few of the relevant shafts have a grid connection. Furthermore, the mountings have no drive units. Due to the costs IWB dispensed with the exchange of the mountings and effort of underground engineering. The solution consists of wireless valve actuators of the 3 S actuators. The first 3 S actuators are put into operation in November 2012.

  20. Physiological water model development

    Science.gov (United States)

    Doty, Susan

    1993-01-01

    The water of the human body can be categorized as existing in two main compartments: intracellular water and extracellular water. The intracellular water consists of all the water within the cells and constitutes over half of the total body water. Since red blood cells are surrounded by plasma, and all other cells are surrounded by interstitial fluid, the intracellular compartment has been subdivided to represent these two cell types. The extracellular water, which includes all of the fluid outside of the cells, can be further subdivided into compartments which represent the interstitial fluid, circulating blood plasma, lymph, and transcellular water. The interstitial fluid surrounds cells outside of the vascular system whereas plasma is contained within the blood vessels. Avascular tissues such as dense connective tissue and cartilage contain interstitial water which slowly equilibrates with tracers used to determine extracellular fluid volume. For this reason, additional compartments are sometimes used to represent these avascular tissues. The average size of each compartment, in terms of percent body weight, has been determined for adult males and females. These compartments and the forces which cause flow between them are presented. The kidneys, a main compartment, receive about 25 percent of the cardiac output and filters out a fluid similar to plasma. The composition of this filtered fluid changes as it flows through the kidney tubules since compounds are continually being secreted and reabsorbed. Through this mechanism, the kidneys eliminate wastes while conserving body water, electrolytes, and metabolites. Since sodium accounts for over 90 percent of the cations in the extracellular fluid, and the number of cations is balanced by the number of anions, considering the renal handling sodium and water only should sufficiently describe the relationship between the plasma compartment and kidneys. A kidney function model is presented which has been adapted from a

  1. Regional long-term model of radioactivity dispersion and fate in the Northwestern Pacific and adjacent seas: application to the Fukushima Dai-ichi accident

    International Nuclear Information System (INIS)

    The compartment model POSEIDON-R was modified and applied to the Northwestern Pacific and adjacent seas to simulate the transport and fate of radioactivity in the period 1945–2010, and to perform a radiological assessment on the releases of radioactivity due to the Fukushima Dai-ichi accident for the period 2011–2040. The model predicts the dispersion of radioactivity in the water column and in sediments, the transfer of radionuclides throughout the marine food web, and subsequent doses to humans due to the consumption of marine products. A generic predictive dynamic food-chain model is used instead of the biological concentration factor (BCF) approach. The radionuclide uptake model for fish has as a central feature the accumulation of radionuclides in the target tissue. The three layer structure of the water column makes it possible to describe the vertical structure of radioactivity in deep waters. In total 175 compartments cover the Northwestern Pacific, the East China and Yellow Seas and the East/Japan Sea. The model was validated from 137Cs data for the period 1945–2010. Calculated concentrations of 137Cs in water, bottom sediments and marine organisms in the coastal compartment, before and after the accident, are in close agreement with measurements from the Japanese agencies. The agreement for water is achieved when an additional continuous flux of 3.6 TBq y−1 is used for underground leakage of contaminated water from the Fukushima Dai-ichi NPP, during the three years following the accident. The dynamic food web model predicts that due to the delay of the transfer throughout the food web, the concentration of 137Cs for piscivorous fishes returns to background level only in 2016. For the year 2011, the calculated individual dose rate for Fukushima Prefecture due to consumption of fishery products is 3.6 μSv y−1. Following the Fukushima Dai-ichi accident the collective dose due to ingestion of marine products for Japan increased in 2011 by a factor

  2. 气象因素对电力安全事故影响的模型%Research on Impact Model of Meteorological Factors on the Power Accidents

    Institute of Scientific and Technical Information of China (English)

    李彦斌; 韩颖; 张嵘; 李彦国

    2013-01-01

      以我国南方某地区为例,收集了该地区连续48个月的电力事故数据及其对应的15个气象要素资料。首先用因子分析法消除了15个气象要素的多重共线性,提取了温度因子、降水因子、湿度因子、风力因子4类主要因素,其次应用Logistic回归建立了气象因素对电力事故的影响模型。模型探究了气象条件与电力事故的内在联系,并用2010年的检验样本验证了模型拟合的准确性,为电力事故预警机制的建立进行了积极的探讨。%Taking a certain region in South China for example, the power system accident data in continuous 48 months of this region and corresponding data of 15 meteorological factors are collected. Firstly, the multicollinearity of the 15 meteorological factors are eliminated by factor analysis, and four kinds of main factors such as temperature factor, precipitation factor, humidity factor and wind power factor are extracted;secondly, an impact model of meteorological factors on power system accidents is established by Logistic regression. Using the established impact model the internal relations between meteorological conditions and power system accidents are explored, and the fitting accuracy of the proposed model is validated by test samples in 2010. The proposed impact model is available for reference to the establishment of early warning mechanism for power system accidents.

  3. Econometric models for biohydrogen development.

    Science.gov (United States)

    Lee, Duu-Hwa; Lee, Duu-Jong; Veziroglu, Ayfer

    2011-09-01

    Biohydrogen is considered as an attractive clean energy source due to its high energy content and environmental-friendly conversion. Analyzing various economic scenarios can help decision makers to optimize development strategies for the biohydrogen sector. This study surveys econometric models of biohydrogen development, including input-out models, life-cycle assessment approach, computable general equilibrium models, linear programming models and impact pathway approach. Fundamentals of each model were briefly reviewed to highlight their advantages and disadvantages. The input-output model and the simplified economic input-output life-cycle assessment model proved most suitable for economic analysis of biohydrogen energy development. A sample analysis using input-output model for forecasting biohydrogen development in the United States is given.