WorldWideScience

Sample records for accident prediction models

  1. Traffic Accident Prediction Model Implementation in Traffic Safety Management

    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...

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

    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

  3. Predictive accident modeling approach inrelation to workover systems

    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...

  4. Combined Prediction Model of Death Toll for Road Traffic Accidents Based on Independent and Dependent Variables

    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.

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

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

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

    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...

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

    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

  8. Debris interactions in reactor vessel lower plena during a severe accident. I. Predictive model

    For pt.II see ibid., p.165-78, 1996. An integral predictive physico-numerical model has been developed to understand and interpret debris interactions in the reactor vessel plenum such as those which took place in the TMI-2 accident. The model represents the extent of debris jet disintegration by a jet-water entrainment model which can result in two types of debris configurations. One is particulated debris which eventually quenches in the water as a result of the entrainment process. The remainder of the debris penetrates to the bottom of the lower plenum and collects as a continuous layer. Each is treated as a separate region and has governing principles for its behavior. The potential for creating gap (contact) resistance and boiling heat removal is considered for heat transfer between the debris bed, the reactor vessel and steel structures and, most importantly, the vessel-to-crust gap water. The proposed in-vessel cooling mechanism due to material creep and water ingression into the expanding gap between the core debris and the vessel wall was found to explain the non-failure of the TMI-2 vessel in the course of the accident. The particulate debris bed is a mixture of metal and oxide, which is distributed as individual spherical particles of sizes determined at the time of entrainment. Energy is received from the continuum bed below by radiation and convection. The continuum debris bed is described by the crust behavior with the heat flux to the crust given by the natural convection correlations relating the Nusselt and Rayleigh numbers for the central region of debris. Using these governing principles, the rate laws for heat and mass transfer are formulated for each type of debris condition in the lower plenum

  9. Survey of accidents in suburban Tehran and the prediction of future events based on a time-series model

    Heidar Teymuri, Ghulam; Bahmani, Rahman; Asghari, Mehdi; Madrese, Elham; Rahmani, Abdolrasoul; Abbasinia, Marzieh; Ahmadnezhad, Iman; Samavati, Mehdi

    2014-01-01

    Background: Car accidents are currently a social issue globally because they result in the deaths of many people. The aim of this study was to examine traffic accidents in suburban Tehran and forecast the number of future accidents using a time-series model. Methods: The sample population of this cross-sectional study was all traffic accidents that caused death and physical injuries in suburban Tehran in 2010 and 2011, as registered by the Tehran Emergency Section. In the present study, Minit...

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

    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.

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

    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.

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

    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.

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

    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.

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

    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

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

    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.

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

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

  17. Review of models applicable to accident aerosols

    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

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

    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...

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

    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

  20. Correspondence model of occupational accidents

    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

  1. Do Cognitive Models Help in Predicting the Severity of Posttraumatic Stress Disorder, Phobia, and Depression after Motor Vehicle Accidents? A Prospective Longitudinal Study

    Ehring, Thomas; Ehlers, Anke; Glucksman, Edward

    2008-01-01

    The study investigated the power of theoretically derived cognitive variables to predict posttraumatic stress disorder (PTSD), travel phobia, and depression following injury in a motor vehicle accident (MVA). MVA survivors (N = 147) were assessed at the emergency department on the day of their accident and 2 weeks, 1 month, 3 months, and 6 months…

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

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

    1997-02-01

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

  3. Injury risk prediction for traffic accidents in Porto Alegre/RS, Brazil

    Perone, Christian S.

    2015-01-01

    This study describes the experimental application of Machine Learning techniques to build prediction models that can assess the injury risk associated with traffic accidents. This work uses an freely available data set of traffic accident records that took place in the city of Porto Alegre/RS (Brazil) during the year of 2013. This study also provides an analysis of the most important attributes of a traffic accident that could produce an outcome of injury to the people involved in the accident.

  4. Prediction of road accidents: A Bayesian hierarchical approach

    Deublein, Markus; Schubert, Matthias; Adey, Bryan T.;

    2013-01-01

    -lognormal regression analysis taking into account correlations amongst multiple dependent model response variables and effects of discrete accident count data e.g. over-dispersion, and (3) Bayesian inference algorithms, which are applied by means of data mining techniques supported by Bayesian Probabilistic Networks...... in order to represent non-linearity between risk indicating and model response variables, as well as different types of uncertainties which might be present in the development of the specific models.Prior Bayesian Probabilistic Networks are first established by means of multivariate regression analysis...... of the observed frequencies of the model response variables, e.g. the occurrence of an accident, and observed values of the risk indicating variables, e.g. degree of road curvature. Subsequently, parameter learning is done using updating algorithms, to determine the posterior predictive probability distributions...

  5. Do cognitive models help in predicting the severity of posttraumatic stress disorder, phobia and depression after motor vehicle accidents? A prospective longitudinal study

    Ehring, T.; Ehlers, A; Glucksman, E.

    2008-01-01

    The study investigated the power of theoretically derived cognitive variables to predict posttraumatic stress disorder (PTSD), travel phobia, and depression following injury in a motor vehicle accident (MVA). MVA survivors (N = 147) were assessed at the emergency department on the day of their accident and 2 weeks, 1 month, 3 months, and 6 months later. Diagnoses were established with the Structured Clinical Interview for DSM–IV. Predictors included initial symptom severities; variables estab...

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

    Levin, A.E.; Griffith, P.

    1980-04-01

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

  7. Modeling accident frequency in Denmark for improving road safety

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

    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 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...... concerning police recorded accidents, link characteristics of the road network, traffic volumes from the national transport models are merged to estimate the model. Spatial correlation between road sections is taken into account for correcting for unobserved correlation between contiguous locations....

  8. Do cognitive models help in predicting the severity of posttraumatic stress disorder, phobia and depression after motor vehicle accidents? A prospective longitudinal study

    T. Ehring; A. Ehlers; E. Glucksman

    2008-01-01

    The study investigated the power of theoretically derived cognitive variables to predict posttraumatic stress disorder (PTSD), travel phobia, and depression following injury in a motor vehicle accident (MVA). MVA survivors (N 147) were assessed at the emergency department on the day of their acciden

  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

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

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

    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.

  12. Modeling accidents for prioritizing prevention

    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

  13. An exploration of the utility of mathematical modeling predicting fatigue from sleep/wake history and circadian phase applied in accident analysis and prevention: the crash of Comair Flight 5191.

    Pruchnicki, Shawn A; Wu, Lora J; Belenky, Gregory

    2011-05-01

    On 27 August 2006 at 0606 eastern daylight time (EDT) at Bluegrass Airport in Lexington, KY (LEX), the flight crew of Comair Flight 5191 inadvertently attempted to take off from a general aviation runway too short for their aircraft. The aircraft crashed killing 49 of the 50 people on board. To better understand this accident and to aid in preventing similar accidents, we applied mathematical modeling predicting fatigue-related degradation in performance for the Air Traffic Controller on-duty at the time of the crash. To provide the necessary input to the model, we attempted to estimate circadian phase and sleep/wake histories for the Captain, First Officer, and Air Traffic Controller. We were able to estimate with confidence the circadian phase for each. We were able to estimate with confidence the sleep/wake history for the Air Traffic Controller, but unable to do this for the Captain and First Officer. Using the sleep/wake history estimates for the Air Traffic Controller as input, the mathematical modeling predicted moderate fatigue-related performance degradation at the time of the crash. This prediction was supported by the presence of what appeared to be fatigue-related behaviors in the Air Traffic Controller during the 30 min prior to and in the minutes after the crash. Our modeling results do not definitively establish fatigue in the Air Traffic Controller as a cause of the accident, rather they suggest that had he been less fatigued he might have detected Comair Flight 5191's lining up on the wrong runway. We were not able to perform a similar analysis for the Captain and First Officer because we were not able to estimate with confidence their sleep/wake histories. Our estimates of sleep/wake history and circadian rhythm phase for the Air Traffic Controller might generalize to other air traffic controllers and to flight crew operating in the early morning hours at LEX. Relative to other times of day, the modeling results suggest an elevated risk of fatigue

  14. China's coal mine accident statistics analysis and one million tons mortality prediction

    Qiao Tong

    2016-03-01

    Full Text Available In order to study the general rule of coal mine accidents in China in recent years, the data of coal mine accident in 2011-2015 is analyzed. The mathematical statistics method is used to analyze the occurrence year, type, season and area of the accident. The results of analysis shows that the coal mine accident has been reduced year by year, and the frequency of gas explosion is the highest. The frequency and the number of deaths in the second quarter of the year are the highest; Guizhou province, Hunan province, Yunnan province and Heilongjiang province are the accident prone provinces. GM (1, 1 dynamic prediction model is used to model and forecast the future million tons mortality in China. The forecast results show that the coal mine's million tons mortality rate of China showed a decreasing trend. The forecast results are scientific and reliable, and it is of great significance to the safety management of coal mine.

  15. The observed and predicted health effects of the Chernobyl accident

    Due to poor design, operator error and the absence of an established Safety Culture, the worst accident in the history of nuclear power involving the Unit 4 RMBK reactor occurred at Chernobyl in the Ukraine in the early morning of 26 April 1986. This accident led to the contamination of large tracts of forest and agricultural land (in the former Soviet Union) and the evacuation of a large number of people. Thirty-one people died at the time of the accident or shortly afterwards, and 203 people were treated for the Acute Radiation Syndrome. From about 1990 a significant increase in the number of childhood thyroid cancers has been noted in Belarus and Ukraine. Because of the social, political and economic situation in the Soviet Union soon after the accident, the anxiety and stress induced in the general population has been enhanced to the point where it may well be the single most important indirect health effect of the accident. Contamination outside the former Soviet Union was largely confined to Europe, where it was extremely patchy and variable. Contamination in the rest of the Northern Hemisphere was insignificant. The health effects in the General Population in the Contaminated Regions in the former USSR and Europe, are predicted to be low and not discernible. However, there may be subgroups within, for example, the Liquidators, which if they can be identified and followed, may show adverse health effects. Health effects in the rest of the Northern Hemisphere will be inconsequential. (author) 38 refs., 1 tab., 1 fig

  16. Validation and verification of accident consequence assessment models

    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)

  17. Risk horoscopes: Predicting the number and type of serious occupational accidents in The Netherlands for sectors and jobs

    The risk of a serious occupational accident per hour exposure was calculated in a project to develop an occupational risk model in the Netherlands (WebORCA). To obtain risk rates, the numbers of victims of serious occupational accidents investigated by the Dutch Labour inspectorate 1998–Feb 2004 were divided by the number of hours exposure for each of 64 different types of hazards, such as contact with moving parts of machines and falls from various types of height. The exposures to the occupational accident hazards were calculated from a survey of a panel of 30,000 from the Dutch working population. Sixty risk rates were then used to predict serious accidents for activity sectors and jobs in the Netherlands where exposures to the hazards for sectors or jobs could be estimated from the survey. Such predictions have been called “horoscopes” because the idea is to provide a quick look-up of predicted accidents for a particular sector or job. Predictions compared favourably with actual data. It is concluded that predictive data can help provide information about accidents in cases where there is a lack of data, such as for smaller sub groups of the working population. - Highlights: • Dutch occupational accident risk rates and yearly exposures for 60 hazards are given. • Risks rates are based on the 1% most serious accidents 1998–Feb 2004. • Risk rates are used to predict serious accident risks in jobs and sectors. • Predictions (“risk horoscopes”) give a good match with actual accidents. • Risk horoscopes can help worker groups identify most important accident risks

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

    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

    The conventional PSA techniques cannot adequately evaluate all events. The conventional PSA models usually focus on single internal events such as DBAs, the external hazards such as fire, seismic. However, the Fukushima accident of Japan in 2011 reveals that very rare event is necessary to be considered in the PSA model to prevent the radioactive release to environment caused by poor treatment based on lack of the information, and to improve the emergency operation procedure. Especially, the results from PSA can be used to decision making for regulators. Moreover, designers can consider the weakness of plant safety based on the quantified results and understand accident sequence based on human actions and system availability. This study is for PSA modeling of combined accidents including total loss of feedwater (TLOFW) accident. The TLOFW accident is a representative accident involving the failure of cooling through secondary side. If the amount of heat transfer is not enough due to the failure of secondary side, the heat will be accumulated to the primary side by continuous core decay heat. Transients with loss of feedwater include total loss of feedwater accident, loss of condenser vacuum accident, and closure of all MSIVs. When residual heat removal by the secondary side is terminated, the safety injection into the RCS with direct primary depressurization would provide alternative heat removal. This operation is called feed and bleed (F and B) operation. Combined accidents including TLOFW accident are very rare event and partially considered in conventional PSA model. Since the necessity of F and B operation is related to plant conditions, the PSA modeling for combined accidents including TLOFW accident is necessary to identify the design and operational vulnerabilities.The PSA is significant to assess the risk of NPPs, and to identify the design and operational vulnerabilities. Even though the combined accident is very rare event, the consequence of combined

  19. Thyroid cancer in children and adolescents of Belarus irradiated as a result of Chernobyl accident: status and prediction

    Thyroid cancer incidence in the human population of Belarus irradiated in childhood for the period passed after the Chernobyl accident is analysed and potential perspectives for development of disease incidence in exposed population during life span. Thyroid cancer cases in children and adolescents of Belarus irradiated due to the Chernobyl accident are predicted using the additive model with modified parameters. Predicted values are shown to be in good agreement with the actual data on thyroid cancer cases in children aged 0-6

  20. Prediction of Reactor Vessel Water Level Using Fuzzy Neural Networks in Severe Accidents due to LOCA

    When the initial events that may lead to the severe accident such as Loss Of Coolant Accident (LOCA) and Steam Generator Tube Rupture (SGTR) occurs at a nuclear power plant, it is most important to check the status of the plant conditions by observing the safety-related parameters such as neutron flux, pressurizer pressure, steam generator pressure and water level. In this paper, we propose a method of predicting the water level of coolant in the reactor vessel that directly affect the important events such as the exposure of the reactor core and the damage of reactor vessel by using a Fuzzy Neural Network (FNN) method. In addition, the data for verifying a proposed model was obtained by simulating the severe accident scenarios for the OPR1000 nuclear power plant using the MAAP4 code. In this paper, a prediction model was developed for predicting the reactor vessel water level using the FNN method. The proposed FNN model was verified based on the simulation data of OPR1000 by using MAAP4 code. As a result of simulation, we could see that the performance of the proposed FNN model is quite satisfactory but some large errors are observed occasionally. If the proposed FNN model is optimized by using a variety of data, it is possible to predict the reactor vessel water level exactly

  1. Modeling secondary accidents identified by traffic shock waves.

    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

  2. 基于改进模糊数量化理论的事故微观预测模型%Micro Prediction Model of Traffic Accident Based on Improved Fuzzy Quantitative Theory

    秦利燕; 秦玉权; 邵春福

    2012-01-01

    针对道路交通事故发生的随机性及影响事故发生的因素很多,而且影响因素中既有定量因素又有定性因素的情况.首先分析了道路交通事故事故4项指标和事故率,确定事故率作为微观预测目标;然后从人-车-路组成的系统观点出发对事故因素分析,选取驾驶员的驾龄、车道数、平曲线半径、纵坡度、路面情况、路口路段类型、道路宽度和交通流量等变量作为主要影响因素,其中前7项作为定性影响因素,交通流量作为定量影响因素,各定性因素下分为若干类目;最后在数量化理论的基础之上建立了改进的模糊道路交通事故微观预测模型.该模型以某国道466.678~559.792 km段作为算例进行计算,计算结果表明:三枝交叉口对事故影响最大,针对该路段提出具体的道路整改意见.%The random feature of traffic accidents with multiple influencing factors includes qualitative and quantitative ones. The accident rate was selected as the microscopic prediction objective after analyzing 4 traffic accident indexes and accident rates. Then several factors, including driving years, number of lanes, radius of horizontal curve, longitudinal grade, road surface status, types of intersection and road section, width of road surface and traffic volume were selected as major influencing factors for analyzing the influencing factors from a systematic perspective with the combination of human, vehicle and road. Among these factors, traffic volume was quantative factor, and the other were qualitative factors which were divided into several categories respectively. An improved fuzzy microscopic model for predicting traffic accidents was established based on quantitative theory. To verify the model, the data of road accidents and a 466. 678 -559. 792 km section on certain national highway was taken for example calculation. The calculation result indicates that three-way intersection has the greatest

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

    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

  4. Safety Performance Improvement for Nuclear Power Plants Using THOMAS and Accident Prediction Function

    The environments of nuclear industry are changed by incoming of digital technology. Until now, the nuclear power plant was adhering to analogue system in the large part of system. However the reliability of digital technology is increased, the adopting of digital technology is accelerated in the nuclear industry. It is not exception of the part of monitoring system. Digital based thermal hydraulics online monitoring advisory system of nuclear power plant, as called THOMAS, which is can be developed economically compared with existing monitoring system is used for the decision making tool in the accident condition. We selected the Ulchin 3 and 4 units which is the type of Korea Standard Nuclear Power Plant(KSNP) as reference plant. For nuclear power plants, EOPs (Emergency Operating Procedures) help operators to diagnose and analyze accidents. But it is very difficult that operators diagnose and analyze similar accidents with EOPs in a given short time. There are also possibilities to follow wrong procedures due to complex and extensive procedures. Therefore, it is important to develop a methodology for diagnosing accidents in a short time and reduction of human errors that made by complex signals and indicators. THOMAS has a function of decision making using influence diagram logic. The influence diagram logic is based on total probability and Bayesian theory. And also the accident modeling is based on emergency operating procedure(EOP). The final goal of this system is, in the accident situation, to present the success path to the operator for the recovery of system. In this paper, at first, we will deal briefly with total system of THOMAS. And then 3D visualized window and accident prediction function will be introduced in detail

  5. Development of accident diagnosis and prediction system for research reactor

    A pilot system of early fault detection expert system has been developed. The early fault detection expert system is one of subsystems in the accident diagnosis and prediction system for the research reactor JRR-3 in JAERI. Functions of the pilot system are to detect deviations of process parameters from the steady state in the early stage of the transient and, if possible, to provide procedures to operators to avoid scram actuation. The reactor accident diagnosis system, DISKET, which had been developed in JAERI, was applied for developing the pilot system by extending functions as follows. (1) A frame structure has been introduced to a part of the knowledge base of DISKET in order to infer efficiently. (2) Numerical equation has been introduced to rule representation in order to calculate numerical value for rules. The pilot system was tested against some simulated transients to validate the effectiveness of the extension mentioned above as well as the performance of the system. This report describes development of the pilot system and the results of the test. (author)

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

    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. MELCOR modeling of Fukushima unit 2 accident

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

    2014-12-15

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

  8. MELCOR modeling of Fukushima unit 2 accident

    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.

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

    Hesamaldin Razi

    2013-01-01

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

  10. HTR fuel: prediction of fission product release in accidents

    The basic fuel unit of the HTR is the coated particle of about 1 mm diameter. An oxidic fuel kernel is surrounded by a low density buffer layer and a silicon carbide coating sandwiched between high density pyrocarbon coatings. The total release of fission products during accidents is determined not only by the transient-induced and the irradiation-induced failure of the coatings, but also by the levels of manufacturing defects and the level of heavy metal contamination in the fuel matrix material. Modern coated fuel particles are designed so that the fission gas pressure-induced stress in the SiC coating remains small relative to the strength of the SiC even under full design burnup conditions. Therefore the pressure vessel failure of the particles is insignificant both in normal operations and in accidents. Silicon carbide thermal decomposition becomes the dominant failure mode as temperatures exceed 2000 deg. C. Interaction of fission products with silicon carbide leading to SiC corrosion is the dominant failure mechanism below 2000 deg. C. Laboratory simulations of HTR transients have usually measured the release of Cs 137 and Kr 85 as indicators of the coating failure. Once the silicon carbide fails by corrosion or decomposition, Cs 137 is released and is taken as the direct indicator of SiC failure in fuel performance modeling studies. In the case of Kr, an additional delay beyond the Cs release is found due to the time required for Kr to diffuse through the remaining outer pyrocarbon coating. The delay between the SiC failure and gas release is analyzed to yield data on the diffusion coefficient of Kr in pyrocarbon. The present data suggest that, in terms of expected values, the fission product release during a modular reactor system transient to 1600 deg. C is dominated by the manufacturing defects and heavy metal contamination rather than irradiation-induced or transient-induced coating failure. (author)

  11. Optimal predictive model selection

    Barbieri, Maria Maddalena; Berger, James O.

    2004-01-01

    Often the goal of model selection is to choose a model for future prediction, and it is natural to measure the accuracy of a future prediction by squared error loss. Under the Bayesian approach, it is commonly perceived that the optimal predictive model is the model with highest posterior probability, but this is not necessarily the case. In this paper we show that, for selection among normal linear models, the optimal predictive model is often the median probability model, which is defined a...

  12. 广西道路交通事故BP人工神经网络预测模型的建立及效果评价%Establishment and Efficacy Evaluation of BP Neural Network Model for Prediction of Road Traffic Accidents in Guangxi

    刘勇; 杨莉; 彭振仁; 黄开勇

    2013-01-01

    目的 构建广西道路交通事故BP人工神经网络预测模型,为研究广西道路交通事故提供新方法.方法 在分析道路交通事故与人、车、路等因素关系的基础上,选取人口数、客运周转量、民用车辆拥有量和公里里程数作为输入变量,交通事故发生数作为输出变量,应用BP人工神经网络技术,对2010年广西道路交通发生数进行预测.结果 2010年广西交通事故预测数为4 562次,实际发现4 351次,预测值与实际值误差为4.85%,建立的模型拟合效果较好.结论 BP人工神经网络模型适用于广西交通事故数的预测,为交通部门进行交通事故预测研究提供新方法.%Objective To construct the prediction model of road traffic accidents in Guangxi by BP neural network, and to provide a new method for studying road traffic accidents. Methods Based on the analysis of the relation between road traffic accidents and factors,including human,vehicle and road,the predicting model of road traffic accidents, which used population,passenger turnover,number of civilian vehicles and Km mileage as the input neurons and the road traffic accidents as the output neuron, was established by BP neural network to predict the road traffic accidents of Guangxi in 2010. Results The predicted value and actual one in 2010 for the road traffic accidents were 4 562 and 4 351,respectively, and the percentage of error was 4. 85%. The fitting of the model established was more effective. Conclusion The predicting model established by BP neural network is suited for predicting road traffic accidents in Guangxi, and it has provided a new method for traffic department.

  13. Prediction of widespread radionuclide contamination at reactor accidents

    JAERI has developed a real-time prediction system, SPEEDI (System for Prediction of Environmental Emergency Dose Information). The system has recently been extended to WSPEEDI (the worldwide version of SPEEDI) by enlarging the application scope using a supercomputer with a network for worldwide meteorological data as included in the wind field model and the dispersion model. This new system can trace the movement of radionuclides over a large area up to the hemisphere. Participating in the international cooperative work, called ATMES (Atmospheric Transport Model Evaluation Study) and ETEX (European Tracer Experiment) projects in which nontoxic artificial tracer gas was released and the concentration monitored at 168 stations located over 2000 km region, JAERI joined in predicting the evolution of concentration distribution as a function of time. The results with WSPEEDI were compared and some future modification of the system is described. (S. Ohno)

  14. Predicting material release during a nuclear reactor accident

    KONINGS Rudy; Wiss, Thierry; BENES ONDREJ

    2014-01-01

    The accident in the Fukushima Daiichi nuclear power plant that happened four years ago this month, has once more drawn the attention of a broad public to the environmental impact of the release of fission products from nuclear power reactors in the event of an accident in which the reactor core is damaged. So far three such accidents have occurred in the history of civil nuclear power production. In this commentary we will review the state-of-the-art of the knowlegde of the physical and chemi...

  15. The prediction of the LWR plant accident based on the measured plant data

    In case of accident affecting a nuclear reactor, it is essential to anticipate the possible development of the situation to efficiently succeed in emergency response actions, i.e. firstly to be early warned, to get sufficient information on the plant: and as far as possible. The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) project consists in developing a methodology: of expertise to; structure the work of technical teams and to facilitate cross competence communications among EP players and a qualified computer tool that could be commonly used by the European countries to reliably predict source term in case of an accident in a light water reactor, using the information available on the plant. In many accident conditions the team of analysts may be located far away from the plant experiencing the accident and their decision making is based on the on-line plant data transmitted into the crisis centre in an interval of 30 - 600 seconds. The plant condition has to be diagnosed based on this information, In the ASTRID project the plant status diagnostics has been studied for the European reactor types including BWR, PWR and VVER plants. The directly measured plant data may be used for estimations of the break size from the primary system and its locations. The break size prediction may be based on the pressurizer level, reactor vessel level, primary pressure and steam generator level in the case of the steam generator tube rupture. In the ASTRID project the break predictions concept was developed and its validity for different plant types and is presented in the paper, when the plant data has been created with the plant specific thermohydraulic simulation model. The tracking simulator attempts to follow the plant behavior on-line based on the measured plant data for the main process parameters and most important boundary conditions. When the plant state tracking fails, the plant may be experiencing an accident, and the tracking

  16. Principles of medical-hygienic prediction of the after-effects of radiation accident

    Basing on the experience in rendering medical aid to the injured persons after radiation accidents during years 1950-1993, the recommendations are given in respect of the procedure of preparation of the medical prognosis of the after-effects of radiation accident with the help of IAEA scale of severity radiation accident. It is pointed out that the prediction of hygienic after-effects of the radiation accident depends on a number of factors, the applied methods of weighing-out the benefit and harm of the introduced hygienic measures and remains to be the subject of further studies. 9 refs.; 3 tabs

  17. 事故预测GM(1,1)模型的Excel求解%Application of Excel to Solve the G(1,1) Model of Accident Prediction

    李杰

    2013-01-01

    GM(1,1)在事故预测上得到了广泛的运用,而GM(1,1)复杂繁琐的计算对于一线的安全管理人员来说使用起来具有一定的难度.而对于一线安全管理人员来讲,EXCEL进行数据管理和分析相对熟悉.为此,使用EXCEL求解GM(1,1)在一线安全管理人员当中成为可能.该文通过实例对GM(1,1)问题进行了求解,并对计算结果进行了验证,说明使用EX?CEL能够精确的求解GM(1,1,)模型.%GM (1,1) has been widely used in the accident prediction, complicated calculation of GM (1,1) for ordinary security management was very difficult for general safety in terms of the management staff, while excel data management and analysis is relatively familiar. To do this, Use the EXCEL that was possible to solving GM (1,1) in ordinary security management officers. This article gave an example of GM (1, 1) problem and solved, and the results was verified. Results show that EXCEL could accu?rately solve the GM (1,1) model.

  18. Grey-Markov Model for Road Accidents Forecasting

    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.

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

    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)

  20. Visualizing Risk Prediction Models

    Vanya Van Belle; Ben Van Calster

    2015-01-01

    Objective Risk prediction models can assist clinicians in making decisions. To boost the uptake of these models in clinical practice, it is important that end-users understand how the model works and can efficiently communicate its results. We introduce novel methods for interpretable model visualization. Methods The proposed visualization techniques are applied to two prediction models from the Framingham Heart Study for the prediction of intermittent claudication and stroke after atrial fib...

  1. Core/concrete interaction model for full scope simulation of severe accidents

    Nuclear plant training simulators have only recently begun to model severe loss-of-coolant accidents in which molten core material can relocate to the bottom of the reactor vessel, fail the vessel, and migrate to the containment. For those accident sequences in which core debris )corium) can accumulate in direct contact with concrete in the containment, the potential for concrete erosion and its phenomenological consequences must be assessed in order that operator training for severe accidents can be attempted. The core/concrete interaction model presented in this paper was developed for the Westinghouse full scope simulator. It allows for extension of transient simulation to conditions beyond vessel failure, and is intended for real-time operator training for severe accidents on a full scope simulator. The model predictions compare favorably with more detailed MAAP calculations

  2. Road Accident Trends in Africa and Europe

    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. Applying Functional Modeling for Accident Management of Nuclear Power Plant

    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

  4. Modified ensemble Kalman filter for nuclear accident atmospheric dispersion: Prediction improved and source estimated

    Highlights: • A modified ensemble Kalmen filter data assimilation method is proposed. • The method can consider four main uncertain parameters in the puff model. • The prediction of radioactive material atmospheric dispersion is improved. • The source release rate and plume rise height are successfully reconstructed. • It can shorten the time lag in the response of ensemble Kalmen filter. - Abstract: Atmospheric dispersion models play an important role in nuclear power plant accident management. A reliable estimation of radioactive material distribution in short range (about 50 km) is in urgent need for population sheltering and evacuation planning. However, the meteorological data and the source term which greatly influence the accuracy of the atmospheric dispersion models are usually poorly known at the early phase of the emergency. In this study, a modified ensemble Kalman filter data assimilation method in conjunction with a Lagrangian puff-model is proposed to simultaneously improve the model prediction and reconstruct the source terms for short range atmospheric dispersion using the off-site environmental monitoring data. Four main uncertainty parameters are considered: source release rate, plume rise height, wind speed and wind direction. Twin experiments show that the method effectively improves the predicted concentration distribution, and the temporal profiles of source release rate and plume rise height are also successfully reconstructed. Moreover, the time lag in the response of ensemble Kalman filter is shortened. The method proposed here can be a useful tool not only in the nuclear power plant accident emergency management but also in other similar situation where hazardous material is released into the atmosphere

  5. Predictive modeling of complications.

    Osorio, Joseph A; Scheer, Justin K; Ames, Christopher P

    2016-09-01

    Predictive analytic algorithms are designed to identify patterns in the data that allow for accurate predictions without the need for a hypothesis. Therefore, predictive modeling can provide detailed and patient-specific information that can be readily applied when discussing the risks of surgery with a patient. There are few studies using predictive modeling techniques in the adult spine surgery literature. These types of studies represent the beginning of the use of predictive analytics in spine surgery outcomes. We will discuss the advancements in the field of spine surgery with respect to predictive analytics, the controversies surrounding the technique, and the future directions. PMID:27286683

  6. Mathematical models for steam generator accident simulation

    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)

  7. Zephyr - the prediction models

    Nielsen, Torben Skov; Madsen, Henrik; Nielsen, Henrik Aalborg;

    2001-01-01

    utilities as partners and users. The new models are evaluated for five wind farms in Denmark as well as one wind farm in Spain. It is shown that the predictions based on conditional parametric models are superior to the predictions obatined by state-of-the-art parametric models....

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

    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...

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

    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...

  10. Model based detection and reconstruction of road traffic accidents

    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...

  11. Simplified evaluation models for total fission number in a criticality accident

    For handling of nuclear fuel during reprocessing or for design of spent-fuel storage and transportation, one needs to know the scale of maximum credible criticality accidents, i.e., the total fission number so as to know the radiological exposure of working personnel as well as the risk to the public in the event of an accident. Some simplified evaluation models for conservatively predicting the number of total fissions during an accident are derived theoretically using the one-point adiabatic reactivity balance model for the homogeneous and thermogenesis systems, respectively, which are frequently seen in nuclear fuel facilities. These simplified evaluation models are subsequently validated with the transient experiment data and actual accident data published to date from the world nuclear community. Some conventionally used simplified evaluation models of this kind are quoted and compared with the results to show the convenience of the current models, having almost no restrictions in the application for any kind of nuclear fuel, material composition, geometry, and dimension, and thus, ensuring adequate margins for predicting the total fission number at the time of a criticality accident

  12. Simplified evaluation models for total fission number in a criticality accident

    Nomura, Yasushi; Okuno, Hiroshi [Japan Atomic Energy Research Inst., Ibaraki (Japan). Dept. of Fuel Cycle Safety Research

    1995-01-01

    For handling of nuclear fuel during reprocessing or for design of spent-fuel storage and transportation, one needs to know the scale of maximum credible criticality accidents, i.e., the total fission number so as to know the radiological exposure of working personnel as well as the risk to the public in the event of an accident. Some simplified evaluation models for conservatively predicting the number of total fissions during an accident are derived theoretically using the one-point adiabatic reactivity balance model for the homogeneous and thermogenesis systems, respectively, which are frequently seen in nuclear fuel facilities. These simplified evaluation models are subsequently validated with the transient experiment data and actual accident data published to date from the world nuclear community. Some conventionally used simplified evaluation models of this kind are quoted and compared with the results to show the convenience of the current models, having almost no restrictions in the application for any kind of nuclear fuel, material composition, geometry, and dimension, and thus, ensuring adequate margins for predicting the total fission number at the time of a criticality accident.

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

    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.

  14. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  15. Prediction of the reactor vessel water level using fuzzy neural networks in severe accident circumstance of NPPs

    Park, Soon Ho; Kim, Dae Seop; Kim, Jae Hwan; Na, Man Gyun [Dept. of Nuclear Engineering, Chosun University, Gwangju (Korea, Republic of)

    2014-06-15

    Safety-related parameters are very important for confirming the status of a nuclear power plant. In particular, the reactor vessel water level has a direct impact on the safety fortress by confirming reactor core cooling. In this study, the reactor vessel water level under the condition of a severe accident, where the water level could not be measured, was predicted using a fuzzy neural network (FNN). The prediction model was developed using training data, and validated using independent test data. The data was generated from simulations of the optimized power reactor 1000 (OPR1000) using MAAP4 code. The informative data for training the FNN model was selected using the subtractive clustering method. The prediction performance of the reactor vessel water level was quite satisfactory, but a few large errors were occasionally observed. To check the effect of instrument errors, the prediction model was verified using data containing artificially added errors. The developed FNN model was sufficiently accurate to be used to predict the reactor vessel water level in severe accident situations where the integrity of the reactor vessel water level sensor is compromised. Furthermore, if the developed FNN model can be optimized using a variety of data, it should be possible to predict the reactor vessel water level precisely.

  16. Sensitivity analysis in severe accidents semi-mechanistic modeling

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

  17. Pilot study of dynamic Bayesian networks approach for fault diagnostics and accident progression prediction in HTR-PM

    Zhao, Yunfei; Tong, Jiejuan; Zhang, Liguo, E-mail: lgzhang@tsinghua.edu.cn; Zhang, Qin

    2015-09-15

    Highlights: • Dynamic Bayesian network is used to diagnose and predict accident progress in HTR-PM. • Dynamic Bayesian network model of HTR-PM is built based on detailed system analysis. • LOCA Simulations validate the above model even if part monitors are lost or false. - Abstract: The first high-temperature-reactor pebble-bed demonstration module (HTR-PM) is under construction currently in China. At the same time, development of a system that is used to support nuclear emergency response is in progress. The supporting system is expected to complete two tasks. The first one is diagnostics of the fault in the reactor based on abnormal sensor measurements obtained. The second one is prognostic of the accident progression based on sensor measurements obtained and operator actions. Both tasks will provide valuable guidance for emergency staff to take appropriate protective actions. Traditional method for the two tasks relies heavily on expert judgment, and has been proven to be inappropriate in some cases, such as Three Mile Island accident. To better perform the two tasks, dynamic Bayesian networks (DBN) is introduced in this paper and a pilot study based on the approach is carried out. DBN is advantageous in representing complex dynamic systems and taking full consideration of evidences obtained to perform diagnostics and prognostics. Pearl's loopy belief propagation (LBP) algorithm is recommended for diagnostics and prognostics in DBN. The DBN model of HTR-PM is created based on detailed system analysis and accident progression analysis. A small break loss of coolant accident (SBLOCA) is selected to illustrate the application of the DBN model of HTR-PM in fault diagnostics (FD) and accident progression prognostics (APP). Several advantages of DBN approach compared with other techniques are discussed. The pilot study lays the foundation for developing the nuclear emergency response supporting system (NERSS) for HTR-PM.

  18. Pilot study of dynamic Bayesian networks approach for fault diagnostics and accident progression prediction in HTR-PM

    Highlights: • Dynamic Bayesian network is used to diagnose and predict accident progress in HTR-PM. • Dynamic Bayesian network model of HTR-PM is built based on detailed system analysis. • LOCA Simulations validate the above model even if part monitors are lost or false. - Abstract: The first high-temperature-reactor pebble-bed demonstration module (HTR-PM) is under construction currently in China. At the same time, development of a system that is used to support nuclear emergency response is in progress. The supporting system is expected to complete two tasks. The first one is diagnostics of the fault in the reactor based on abnormal sensor measurements obtained. The second one is prognostic of the accident progression based on sensor measurements obtained and operator actions. Both tasks will provide valuable guidance for emergency staff to take appropriate protective actions. Traditional method for the two tasks relies heavily on expert judgment, and has been proven to be inappropriate in some cases, such as Three Mile Island accident. To better perform the two tasks, dynamic Bayesian networks (DBN) is introduced in this paper and a pilot study based on the approach is carried out. DBN is advantageous in representing complex dynamic systems and taking full consideration of evidences obtained to perform diagnostics and prognostics. Pearl's loopy belief propagation (LBP) algorithm is recommended for diagnostics and prognostics in DBN. The DBN model of HTR-PM is created based on detailed system analysis and accident progression analysis. A small break loss of coolant accident (SBLOCA) is selected to illustrate the application of the DBN model of HTR-PM in fault diagnostics (FD) and accident progression prognostics (APP). Several advantages of DBN approach compared with other techniques are discussed. The pilot study lays the foundation for developing the nuclear emergency response supporting system (NERSS) for HTR-PM

  19. Prediction of temperature and fission product release from HTR fuel under accident conditions

    Modern, small High-Temperature Reactors (HTRs) are designed such that maximum accident fuel temperatures remain below 1600degC without active control mechanisms. It has been demonstrated that HTR fuel remains intact and retains all fission products under these maximum accident conditions at least as well as under normal operating conditions. The accident temperature limit has been achieved by a core design with small thermal power and low power density. In the case of a loss-of-coolant accident (LOCA), the decay heat is removed from the core by passive means. The passive core temperature limitation has been demonstrated with a series of LOCA simulation tests with the AVR pebble-bed HTR in Julich, Germany. Here, the maximum core temperatures were measured to be 1080degC in agreement with predictions and, being used for code validation, in agreement with post-test calculations. (J.P.N.)

  20. Accident progression modelling: containment event trees

    Containment Event Trees (CETs) are used to represent the various potential accident progressions following core melt. The EVNTRE code has a sophisticated Monte-Carlo capability. In this paper the small CET approach uses Decompositions Event Trees (DETs) to analyse the issues behind the CET headers and large CET approach (EVNTRE/NUREG-1150) are presented. The equipment survivability impact in CET, source term assignment via grouping of sequences into categories or by use of parametric code, sensitivity studies versus full Monte-Carlo simulation for study of the impact of uncertainties are also discussed

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

    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.

  2. FIRAC, Nuclear Power Plant Fire Accident Model

    1 - Description of program or function: FIRAC predicts fire-induced flows, thermal and material transport, and radioactive and non- radioactive source terms in a ventilation system. It is designed to predict the radioactive and nonradioactive source terms that lead to gas dynamic, material transport, and heat transfer transients. FIRAC's capabilities are directed toward nuclear fuel cycle facilities and the primary release pathway - the ventilation system. However, it is applicable to other facilities and can be used to model other airflow pathways within a structure. The basic material transport capability of FIRAC includes estimates of entrainment, convection, deposition, and filtration of material. The interrelated effects of filter plugging, heat transfer, and gas dynamics are also simulated. A ventilation system model includes elements such as filters, dampers, ducts, and blowers connected at nodal points to form networks. A zone-type compartment fire model is incorporated to simulate fire-induced transients within a facility. 2 - Method of solution: FIRAC solves one-dimensional, lumped-parameter, compressible flow equations by an implicit numerical scheme. The lumped-parameter method is the basic formulation that describes the gas dynamics system. No spatial distribution of parameters is considered in this approach, but an effect of spatial distribution can be approximated by noding. Network theory, using the lumped-parameter method, includes a number of system elements, called branches, joined at certain points, called nodes. Ventilation system components that exhibit flow resistance and inertia, such as dampers, ducts, valves, and filters, and those that exhibit flow potential, such as blowers, are located within the branches of the system. The connection points of branches are nodes for components that have finite volumes, such as rooms, gloveboxes, and plenums, and for boundaries where the volume is practically infinite. All internal nodes, therefore

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

    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. A NEW HAZARD EVALUATION PROCEDURE FOR PREDICTING RISK FACTORS OF OCCUPATIONAL ACCIDENTS

    Hüseyin CEYLAN

    2013-05-01

    Full Text Available With annual average of 73,937 occupational accidents and 1,152 deaths, Turkey still faces an important problem. The country exercises one of the lowest performances in job safety among the European Union countries. Developments in technology increased the importance of safety management in industry. These improvements also resulted in a requirement of more investment and assignment on human in work systems. This situation increases the importance of forecasting the possible accidents that workers can face and preventing the accidents by taking necessary precautions. In this study a prognostic model has been developed to forecast the occupational accidents in coming periods at the departments of the facilities in hazardous work systems. The validity of the proposed model has been proved by implementing it into practice in hazardous work systems in the manufacturing industry.

  5. STRATEGY PATTERNS PREDICTION MODEL

    Aram Baruch Gonzalez Perez; Jorge Adolfo Ramirez Uresti

    2014-01-01

    Multi-agent systems are broadly known for being able to simulate real-life situations which require the interaction and cooperation of individuals. Opponent modeling can be used along with multi-agent systems to model complex situations such as competitions like soccer games. In this study, a model for predicting opponent moves based on their target is presented. The model is composed by an offline step (learning phase) and an online one (execution phase). The offline step gets and analyses p...

  6. A MELCOR model of Fukushima Daiichi Unit 3 accident

    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

  7. A MELCOR model of Fukushima Daiichi Unit 3 accident

    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.

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

    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.

  9. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment.

    Thoerring, H.; Liland, A.

    2010-12-15

    This report deals with the environmental consequences in Norway after a hypothetical accident at Sellafield. The investigation is limited to the terrestrial environment, and focus on animals grazing natural pastures, plus wild berries and fungi. Only 137Cs is considered. The predicted consequences are severe, in particular for mutton and goat milk production. (Author)

  10. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment

    This report deals with the environmental consequences in Norway after a hypothetical accident at Sellafield. The investigation is limited to the terrestrial environment, and focus on animals grazing natural pastures, plus wild berries and fungi. Only 137Cs is considered. The predicted consequences are severe - in particular for mutton and goat milk production. (Author)

  11. Application of GIS in prediction and assessment system of off-site accident consequence for NPP

    The assessment and prediction software system of off-site accident consequence for Guangdong Nuclear Power Plant (GNARD2.0) is a GIS-based software system. The spatial analysis of radioactive materials and doses with geographic information is available in this system. The structure and functions of the GNARD system and the method of applying ArcView GIS are presented

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

    Tusheva, Polina

    2012-03-09

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

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

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

  14. Predictive models in urology.

    Cestari, Andrea

    2013-01-01

    Predictive modeling is emerging as an important knowledge-based technology in healthcare. The interest in the use of predictive modeling reflects advances on different fronts such as the availability of health information from increasingly complex databases and electronic health records, a better understanding of causal or statistical predictors of health, disease processes and multifactorial models of ill-health and developments in nonlinear computer models using artificial intelligence or neural networks. These new computer-based forms of modeling are increasingly able to establish technical credibility in clinical contexts. The current state of knowledge is still quite young in understanding the likely future direction of how this so-called 'machine intelligence' will evolve and therefore how current relatively sophisticated predictive models will evolve in response to improvements in technology, which is advancing along a wide front. Predictive models in urology are gaining progressive popularity not only for academic and scientific purposes but also into the clinical practice with the introduction of several nomograms dealing with the main fields of onco-urology. PMID:23423686

  15. Modeling alternative clad behavior for accident tolerant systems

    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)

  16. STRATEGY PATTERNS PREDICTION MODEL

    Aram Baruch Gonzalez Perez

    2014-01-01

    Full Text Available Multi-agent systems are broadly known for being able to simulate real-life situations which require the interaction and cooperation of individuals. Opponent modeling can be used along with multi-agent systems to model complex situations such as competitions like soccer games. In this study, a model for predicting opponent moves based on their target is presented. The model is composed by an offline step (learning phase and an online one (execution phase. The offline step gets and analyses previous experiences while the online step uses the data generated by offline analysis to predict opponent moves. This model is illustrated by an experiment with the RoboCup 2D Soccer Simulator. The proposed model was tested using 22 games to create the knowledge base and getting an accuracy rate over 80%.

  17. Consequences in Norway after a hypothetical accident at Sellafield - Predicted impacts on the environment

    This report describes the possible environmental consequences for Norway due to a hypothetical accident at the Sellafield complex in the UK. The scenario considered involves an explosion and fire at the B215 facility resulting in a 1 % release of the total HAL (Highly Active liquor) inventory of radioactive waste with a subsequent air transport and deposition in Norway. Air transport modelling is based on real meteorological data from October 2008 with wind direction towards Norway and heavy precipitation. This weather is considered to be quite representative as typical seasonal weather. Based on this weather scenario, the estimated fallout in Norway will be ∼ 17 P Bq of caesium-137 which is 7 times higher than the fallout from the Chernobyl accident. The modelled radioactive contamination is linked with data on transfer to the food chain and statistics on production and hunting to assess the consequences for foodstuffs. The investigation has been limited to the terrestrial environment, focussing on wild berries, fungi, and animals grazing unimproved pastures (i.e. various types of game, reindeer, sheep and goats). The predicted consequences are severe - especially in connection to sheep and goat production. Up to 80 % of the lambs in Norway could be exceeding the food intervention levels for radiocaesium the first years after the fallout, with 30-40 % likely to be above for many years. There will, consequently, be a need for extensive countermeasures in large areas for years or even decades involving several hundred thousand animals each year. Large consequences are also expected for reindeer husbandry - the first year in particular due to the time of fallout which is just prior to winter slaughter. The consequences will be most sever for reindeer herding in middle and southern parts of Norway, but problems may reach as far north as Finnmark where we find the majority of Norwegian reindeer production. The consequences for game will mostly depend on the regional

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

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

  19. Prediction model Perla

    Prediction model Perla presents one of a tool for an evaluation of a stream ecological status. It enables a comparing with a standard. The standard is formed by a dataset of sites from all area of the Czech Republic. The sites were influenced by a human activity as few as possible. 8 variables were used for prediction (distance from source, elevation, stream width and depth, slope, substrate roughness, longitude and latitude. All of them were statistically important for benthic communities. Results do not response ecoregions, but rather stream size (type). B (EQItaxonu), EQISi, EQIASPT a EQIH appears applicable for assessment using the prediction model and for natural and human stress differentiating. Limiting values of the indices for good ecological status are suggested. On the contrary, using of EQIEPT a EQIekoprof indices would be possible only with difficulties. (authors)

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

    Khater Hany; Abu-El-Maty Talal; El-Morshdy El-Din Salah

    2006-01-01

    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...

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

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

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

    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.

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

    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.

  4. Advanced accident sequence precursor analysis level 2 models

    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.

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

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

    2011-06-01

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

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

    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 ...

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

    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

  8. Development and application of random walk model of atmospheric diffusion in emergency response of nuclear accidents

    Plume concentration prediction is one of the main contents of radioactive consequence assessment for early emergency to nuclear accidents. This paper describes random characteristics of atmospheric diffusion itself, introduces random walk model of atmospheric diffusion (Random Walk), and compare with Lagrangian puff model (RIMPUFF) in the nuclear emergency decision support system (RODOS) developed by European Community for verification. The results show the concentrations calculated by the two models are quite close except that 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 atmospheric diffusion module of the nuclear emergency decision support system. (authors)

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

    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

  10. PSA modeling of long-term accident sequences

    In the traditional Level 1 PSA, the long term of the accident sequences is usually taken into account in a simplified manner. For example, some of the mitigations which are needed at long term are taken into account in the PSA, but the analysis and the associated failures probabilities quantification are estimated based on generic assessments. In the context of the extension of PSA scope to include the external hazards, in France, both operator (EDF) and IRSN work for the improvement of methods to better take into account in the PSA the long term of accident sequences induced by initiators which affect the whole site containing several nuclear installations (reactors, fuel pools, ...). This is an essential prerequisite for the development of external hazards PSA. It has to be noted that in the French PSA, even before Fukushima, this type of accident sequences was already taken into account, many insight being used, as complementary information, to enhance the safety level of the plants. The recent French and international operating experience is an opportunity for tuning the actual PSA methods for long term accident sequences modeling. The paper presents the main results of the ongoing efforts in this area. (author)

  11. Prediction of failure of highly irradiated Zircaloy clad tubes under reactivity initiated accidents

    This paper deals with failure of irradiated Zircaloy tubes under the heat-up stage of a reactivity initiated accident (RIA). More precisely, by use of a model for plastic strain localization and necking failure, we theoretically analyse the effects of local surface defects on clad ductility and survivability under RIA. The results show that even very shallow surface defects, e.g. arising from a non-uniform or partially spilled oxide layer, have a strong limiting effect on clad ductility. Moreover, in presence of surface defects, the ability of the clad tube to expand radially without necking failure is found to be extremely sensitive to the stress biaxiality ratio σzz/σθθ, which is here assumed to be in the range from 0 to 1. The results of our analysis are compared with clad ductility data available in literature, and their consequences for clad failure prediction under RIA are discussed. In particular, the results raise serious concerns regarding the applicability of failure criteria, which are based on clad strain energy density. These criteria do not capture the observed sensitivity to stress biaxiality on clad failure propensity. (author)

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

    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

  13. A model national emergency plan for radiological accidents

    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

  14. A model national emergency response plan for radiological accidents

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

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

    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.

  16. Severe accident development modeling and evaluation for CANDU

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

    2009-09-15

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

  17. Accident consequence assessments with different atmospheric dispersion models

    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.)

  18. Concept and validation studies of the real-time reactor-accident consequences assessment model ECOSYS

    The Chernobyl accident has demonstrated the urgent need for computer programs for real-time assessment of potential radiological consequences of major reactor accidents and for timely recommendations of useful and cost-efficient counter measures. During the past decade the dynamic radioecological program ECOSYS has been developed for nuclear accident consequence assessment with high resolution in space, time and exposure pathways. The Chernobyl reactor accident leading to relatively high contamination of Southern Germany provided excellent conditions for realistic validation studies of concept, sub-models and parameters of ECOSYS. To this purpose more than 7000 low level and in-situ gamma spectroscopy measurements were performed to study experimentally the behaviour of radionuclides in foodchains and in the urban environment and to compare the results to theoretical predictions of ECOSYS. The results show good agreement in the contamination levels of important food stuffs and in external exposure dose rates from a given surface contamination. Improvements were necessary in the assumptions regarding the food consumption habits which changed considerably - and in the functions describing the weathering off from urban and plant surfaces. The results of this validation study and the concept of the improved computerised model, which has subsequently been converted into a real-time code, are discussed in detail

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

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

    2009-06-15

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

  20. Dynamic modelling of radionuclide uptake by marine biota: application to the Fukushima nuclear power plant accident.

    Vives i Batlle, Jordi

    2016-01-01

    The dynamic model D-DAT was developed to study the dynamics of radionuclide uptake and turnover in biota and sediments in the immediate aftermath of the Fukushima accident. This dynamics is determined by the interplay between the residence time of radionuclides in seawater/sediments and the biological half-lives of elimination by the biota. The model calculates time-variable activity concentration of (131)I, (134)Cs, (137)Cs and (90)Sr in seabed sediment, fish, crustaceans, molluscs and macroalgae from surrounding activity concentrations in seawater, with which to derive internal and external dose rates. A central element of the model is the inclusion of dynamic transfer of radionuclides to/from sediments by factorising the depletion of radionuclides adsorbed onto suspended particulates, molecular diffusion, pore water mixing and bioturbation, represented by a simple set of differential equations coupled with the biological uptake/turnover processes. In this way, the model is capable of reproducing activity concentration in sediment more realistically. The model was used to assess the radiological impact of the Fukushima accident on marine biota in the acute phase of the accident. Sediment and biota activity concentrations are within the wide range of actual monitoring data. Activity concentrations in marine biota are thus shown to be better calculated by a dynamic model than with the simpler equilibrium approach based on concentration factors, which tends to overestimate for the acute accident period. Modelled dose rates from external exposure from sediment are also significantly below equilibrium predictions. The model calculations confirm previous studies showing that radioactivity levels in marine biota have been generally below the levels necessary to cause a measurable effect on populations. The model was used in mass-balance mode to calculate total integrated releases of 103, 30 and 3 PBq for (131)I, (137)Cs and (90)Sr, reasonably in line with previous

  1. Candidate Prediction Models and Methods

    Nielsen, Henrik Aalborg; Nielsen, Torben Skov; Madsen, Henrik;

    2005-01-01

    This document lists candidate prediction models for Work Package 3 (WP3) of the PSO-project called ``Intelligent wind power prediction systems'' (FU4101). The main focus is on the models transforming numerical weather predictions into predictions of power production. The document also outlines the...

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

    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

  3. Melanoma risk prediction models

    Nikolić Jelena

    2014-01-01

    Full Text Available Background/Aim. The lack of effective therapy for advanced stages of melanoma emphasizes the importance of preventive measures and screenings of population at risk. Identifying individuals at high risk should allow targeted screenings and follow-up involving those who would benefit most. The aim of this study was to identify most significant factors for melanoma prediction in our population and to create prognostic models for identification and differentiation of individuals at risk. Methods. This case-control study included 697 participants (341 patients and 356 controls that underwent extensive interview and skin examination in order to check risk factors for melanoma. Pairwise univariate statistical comparison was used for the coarse selection of the most significant risk factors. These factors were fed into logistic regression (LR and alternating decision trees (ADT prognostic models that were assessed for their usefulness in identification of patients at risk to develop melanoma. Validation of the LR model was done by Hosmer and Lemeshow test, whereas the ADT was validated by 10-fold cross-validation. The achieved sensitivity, specificity, accuracy and AUC for both models were calculated. The melanoma risk score (MRS based on the outcome of the LR model was presented. Results. The LR model showed that the following risk factors were associated with melanoma: sunbeds (OR = 4.018; 95% CI 1.724- 9.366 for those that sometimes used sunbeds, solar damage of the skin (OR = 8.274; 95% CI 2.661-25.730 for those with severe solar damage, hair color (OR = 3.222; 95% CI 1.984-5.231 for light brown/blond hair, the number of common naevi (over 100 naevi had OR = 3.57; 95% CI 1.427-8.931, the number of dysplastic naevi (from 1 to 10 dysplastic naevi OR was 2.672; 95% CI 1.572-4.540; for more than 10 naevi OR was 6.487; 95%; CI 1.993-21.119, Fitzpatricks phototype and the presence of congenital naevi. Red hair, phototype I and large congenital naevi were

  4. Source term modelling in case of nuclear accidents

    The relative isotopic composition of the nuclides released during a nuclear accidents depends strongly on the implied mechanisms in the failure of fuel elements, safety barriers and accident dynamics. Also, the released fraction depends on the volatility degree and the temperature attaint in the reactor core and the fuel elements during the accident, respectively. At regime operation temperature, when the fuel sheaths are failed the noble gases (Xe and Kr isotopes), the extremely volatile and volatile fission products (I isotopes and Cs, Te and Ru, respectively) are released into the reactor primary circuit. As the temperature increases, other isotopes are released too. Two tables are given presenting a classification of the isotopes in groups of boiling and melting point temperatures, respectively. From the radiologic point of view, evaluation of the impact of the contaminant radioactive release requires consideration of several factors, namely: - activity, half-life, chemical form, biological hazard, geometrical size of the radioactive aerosols, etc. The activity of each isotope at the reactor stack or at the external walls of the reactor building is called source term. The isotopic and combined activity in a point of the environment located at a given distance from the source is evaluated by means of dispersion models starting from the source term. An expression of the activity of a given isotope in terms of its reactor core inventory and the parameters of the safety barriers is presented

  5. Markov Model of Severe Accident Progression and Management

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

    2012-06-25

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

  6. Markov Model of Severe Accident Progression and Management

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

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

    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

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

    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.

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

    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

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

    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

  11. Advanced accident sequence precursor analysis level 1 models

    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.

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

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

  13. Accident sequence modeling: human actions, system response, intelligent decision support

    In Probabilistic Safety Assessment (PSA) of large technological systems, accident sequence modeling represents the synthesis of expert judgement, system modeling, and operational evidence. This book contains the papers that were presented at a two-day Seminar that was held in Munich in August 1987. The aim of this Seminar was to provide a forum for in-depth discussion in a workshop atmosphere of the key elements in the modeling process, such as operator actions and system response, and to assess the possibilities of using such models to design operator decision support systems in the form of expert systems or interactive man computer structures. While this evaluation of the state of the art was done in the context of nuclear power reactor safety, most of the models and ideas advanced by the participants have wide applicability and can be used in safety assessments and reliability enhancement programs for other fields, for example the chemical process and aerospace industries. (author)

  14. A MELCOR model of Fukushima Daiichi Unit 1 accident

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

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

  16. Markov Model of Accident Progression at Fukushima Daiichi

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

    2012-11-11

    On March 11, 2011, a magnitude 9.0 earthquake followed by a tsunami caused loss of offsite power and disabled the emergency diesel generators, leading to a prolonged station blackout at the Fukushima Daiichi site. After successful reactor trip for all operating reactors, the inability to remove decay heat over an extended period led to boil-off of the water inventory and fuel uncovery in Units 1-3. A significant amount of metal-water reaction occurred, as evidenced by the quantities of hydrogen generated that led to hydrogen explosions in the auxiliary buildings of the Units 1 & 3, and in the de-fuelled Unit 4. Although it was assumed that extensive fuel damage, including fuel melting, slumping, and relocation was likely to have occurred in the core of the affected reactors, the status of the fuel, vessel, and drywell was uncertain. To understand the possible evolution of the accident conditions at Fukushima Daiichi, a Markov model of the likely state of one of the reactors was constructed and executed under different assumptions regarding system performance and reliability. The Markov approach was selected for several reasons: It is a probabilistic model that provides flexibility in scenario construction and incorporates time dependence of different model states. It also readily allows for sensitivity and uncertainty analyses of different failure and repair rates of cooling systems. While the analysis was motivated by a need to gain insight on the course of events for the damaged units at Fukushima Daiichi, the work reported here provides a more general analytical basis for studying and evaluating severe accident evolution over extended periods of time. This work was performed at the request of the U.S. Department of Energy to explore 'what-if' scenarios in the immediate aftermath of the accidents.

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

    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

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

    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)

  19. Heat and fluid flow in accident of Fukushima Daiichi Nuclear Power Plant, Unit 1. Accident scenario based on thermodynamic model

    An accident scenario of Fukushima Daiichi Nuclear Power Plant, Unit 1 is analyzed from the data open to the public. Two thermodynamic modes are introduced i.e. a phase equilibrium process model in the reactor pressure vessel (RPV) and an adiabatic model in the pressure containment vessel (PVC). Almost the measured data and observed evidences are explained by the scenario that the isolation condenser was working and a crack at RPV opened at the initial stage of the accident, which is different from TEPCO and the government reports. (author)

  20. Risk forecasting and evaluating model of Environmental pollution accident

    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.

  1. The accident consequence model of the German safety study

    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.)

  2. Global atmospheric dispersion modelling after the Fukushima accident

    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

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

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the food pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 87 imprecisely-known input variables on the following reactor accident consequences are studied: crop growing season dose, crop long-term dose, milk growing season dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, area dependent cost, crop disposal cost, milk disposal cost, condemnation area, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: fraction of cesium deposition on grain fields that is retained on plant surfaces and transferred directly to grain, maximum allowable ground concentrations of Cs-137 and Sr-90 for production of crops, ground concentrations of Cs-134, Cs-137 and I-131 at which the disposal of milk will be initiated due to accidents that occur during the growing season, ground concentrations of Cs-134, I-131 and Sr-90 at which the disposal of crops will be initiated due to accidents that occur during the growing season, rate of depletion of Cs-137 and Sr-90 from the root zone, transfer of Sr-90 from soil to legumes, transfer of Cs-137 from soil to pasture, transfer of cesium from animal feed to meat, and the transfer of cesium, iodine and strontium from animal feed to milk

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

    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.

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

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

    1993-09-01

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

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

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

  7. Specific features of cesium chemistry and physics affecting reactor accident source term predictions

    In the process of assessing remaining uncertainties in predicting the source term of severe reactor accidents, a special investigation is devoted in this report to the case of cesium. The cesium isotopes, especially Cs137 and Cs134, are among those nuclides which could have a major impact on the environment in the event of a release. The processes for release from fuel and retention in the reactor coolant system and the containment are presented. Releases to the atmosphere are also discussed. The intention is to identify and discuss those specific features of cesium chemistry and physics that strongly affect source term predictions. The report has been prepared on contract from the Swedish Nuclear Power Inspectorate as a contribution to the cooperative work within international experts groups of OECD/NEA

  8. ATMOSPHERIC MODELING IN SUPPORT OF A ROADWAY ACCIDENT

    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.

  9. a Simplified Methodology for the Prediction of the Small Break Loss-Of Accident.

    Ward, Leonard William

    1988-12-01

    This thesis describes a complete methodology which has allowed for the development of a faster than real time computer program designed to simulate a small break loss -of-coolant accident in the primary system of a pressurized water reactor. By developing an understanding of the major phenomenon governing the small break LOCA fluid response, the system model representation can be greatly simplified leading to a very fast executing transient system blowdown code. Because of the fast execution times, the CULSETS code, or Columbia University Loss-of-Coolant Accident and System Excursion Transient Simulator code, is ideal for performing parametric studies of Emergency Core Cooling system or assessing the consequences of the many operator actions performed to place the system in a long term cooling mode following a small break LOCA. While the methodology was designed with specific application to the small break loss-of-coolant accident, it can also be used to simulate loss-of-feedwater, steam line breaks, and steam generator tube rupture events. The code is easily adaptable to a personal computer and could also be modified to provide the primary and secondary system responses to supply the required inputs to a simulator for a pressurized water reactor.

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

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

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

    Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the chronic exposure pathways associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 75 imprecisely known input variables on the following reactor accident consequences are studied: crop growing-season dose, crop long-term dose, water ingestion dose, milk growing-season dose, long-term groundshine dose, long-term inhalation dose, total food pathways dose, total ingestion pathways dose, total long-term pathways dose, total latent cancer fatalities, area-dependent cost, crop disposal cost, milk disposal cost, population-dependent cost, total economic cost, condemnation area, condemnation population, crop disposal area and milk disposal area. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: dry deposition velocity, transfer of cesium from animal feed to milk, transfer of cesium from animal feed to meet, ground concentration of Cs-134 at which the disposal of milk products will be initiated, transfer of Sr-90 from soil to legumes, maximum allowable ground concentration of Sr-90 for production of crops, fraction of cesium entering surface water that is consumed in drinking water, groundshine shielding factor, scale factor defining resuspension, dose reduction associated with decontamination, and ground concentration of I-131 at which disposal of crops will be initiated due to accidents that occur during the growing season. Reducing the uncertainty in the preceding variables was found to substantially reduce the uncertainty in the

  12. Aspects of uncertainty analysis in accident consequence modeling

    Mathematical models are frequently used to determine probable dose to man from an accidental release of radionuclides by a nuclear facility. With increased emphasis on the accuracy of these models, the incorporation of uncertainty analysis has become one of the most crucial and sensitive components in evaluating the significance of model predictions. In the present paper, we address three aspects of uncertainty in models used to assess the radiological impact to humans: uncertainties resulting from the natural variability in human biological parameters; the propagation of parameter variability by mathematical models; and comparison of model predictions to observational data

  13. Predictive Modelling Techniques in Radioactive Waste Management

    This paper presents the 'state-of-art' computational modelling techniques AMEC Nuclear has used in radioactive waste management projects. These techniques have been employed to conduct option studies and assessments of radioactive waste packages to justify compliance with the UK and IAEA regulations. An important aspect of a safety case for any packaging is its performance under accident conditions. One of the key principles underlying regulations for performance under normal and accident conditions is that activity release should be low and predictable. This paper addresses the challenge faced by designers and manufacturers to predict behaviour of waste of waste packages has usually been demonstrated by test. Carrying out a full-scale drop test or a fire test of a prototype package with a representative simulant wasteform is time consuming, costly, and can lead to variability in the results. The post-test measurements of release are not straightforward and may be difficult to interpret. Furthermore, these tests are unique for a particular design and cannot be easily applied to other designs. Therefore, predictive modelling based on computational techniques like the finite element analysis (FEA) can be of great benefit. Through examples, the paper examples, the paper explains how assessments of radioactive waste packaging under fire and impact hazards have been conducted to calculate release of radioactive nuclides. The examples include computational modeling to assess free drop and transportation loads on a packaging designed to transportation loads on a packaging designed transport a 50 Te steel pot containing radioactive silicate slag. Methodology used to estimate release fractions from a 500 litre drum following a standard fire assessment is also presented

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

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

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

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

  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

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

    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

  18. Prediction of fission product and aerosol behaviour during a postulated severe accident in a LWR

    Lack of appropriate energy removal causes fuel elements in a reactor core to overheat and may eventually cause core to degrade. Fission products will be emitted from a degraded reactor core. Aerosols are generated when the vapours of various fuel and structural materials reach a cold environment and nucleate. In addition to the fission products release and aerosol generation taking place in the reactor vessel, some more fission products release and aerosol generation will occur when the molten core debris leaves the pressure vessel bottom head and comes in contact with the pedestal concrete floor. Fission products, if they are released to environment from the containment boundary, exert a great danger to public health. A source term is defined as the quantity, timing, and characteristics of the release of radionuclide material to the environment following a postulated severe accident. At PSI a considerable effort hase been spent in investigating and establishing a source term assessment methodology in order to predict the source term for a given Light Water Reactor (LWR) accident scenario. This report introduces the computer programs and the methods associated with the release of the fission products, generation of the aerosols and behaviour of the aerosols in LWR compartments used for a source term assessment analysis at PSI. (author) 4 figs., 5 tabs., 28 refs

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

    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.

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

    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

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

    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.).

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

    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. Modeling and forecasting of accidents at nuclear industrial plants

    The papers on methodology of risk analysis are briefly reviewed. An analysis is performed for relationships between natural and technology-associated accidents. The program of works intended to create a standardization-methodical base of risk analysis at nuclear industrial plants is reported. A number of shortcomings is noted to exist in evaluating nuclear plant safety with the help of commonly used probabilistic criteria of safety. An algorithm of ecological-mathematical monitoring of potentially dangerous objects is suggested. It is pointed out that when developing mathematical models of potentially dangerous object operation not only technological processes, the stochasticity of heat- and mass transfer processes, environmental parameters should be taken into account but social and economical aspects as well

  4. RAPTA-5 code: Modelling behaviour of WWER-type fuel rods in design basis accidents verification calculations

    RAPTA-5 code used for licensing calculations to validate the compliance with the requirements for WWER fuel safety in design basis accidents. The characteristic results are given of design modelling experiments simulating thermomechanical and corrosion behaviour of WWER and PWR fuel rods in LOCA. The results corroborate the adequate predictability of both individual design models and the code as a whole. (author). 14 refs, 12 figs

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

    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...

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

    A new in-vessel fission product release model has been developed and implemented to perform best-estimate calculations of realistic source terms including fuel morphology effects. The proposed bulk mass transfer correlation determines the product of fission product release and equiaxed grain size as a function of the inverse fuel temperature. The model accounts for the fuel-cladding interaction over the temperature range between 770 K and 3000 K in the steam environment. A separate driver has been developed for the in-vessel thermal hydraulic and fission product behavior models that were developed by the Department of Energy for the Modular Accident Analysis Package (MAAP). Calculational results of these models have been compared to the results of the Power Burst Facility Severe Fuel Damage tests. The code predictions utilizing the mass transfer correlation agreed with the experimentally determined fractional release rates during the course of the heatup, power hold, and cooldown phases of the high temperature transients. Compared to such conventional literature correlations as the steam oxidation model and the NUREG-0956 correlation, the mass transfer correlation resulted in lower and less rapid releases in closer agreement with the on-line and grab sample data from the Severe Fuel Damage tests. The proposed mass transfer correlation can be applied for best-estimate calculations of fission products release from the UO2 fuel in both nominal and severe accident conditions. 15 refs., 10 figs., 2 tabs

  7. Prediction models in complex terrain

    Marti, I.; Nielsen, Torben Skov; Madsen, Henrik; Navarro, J.; Barquero, C.G.

    2001-01-01

    The objective of the work is to investigatethe performance of HIRLAM in complex terrain when used as input to energy production forecasting models, and to develop a statistical model to adapt HIRLAM prediction to the wind farm. The features of the terrain, specially the topography, influence the...... performance of HIRLAM in particular with respect to wind predictions. To estimate the performance of the model two spatial resolutions (0,5 Deg. and 0.2 Deg.) and different sets of HIRLAM variables were used to predict wind speed and energy production. The predictions of energy production for the wind farms...... are calculated using on-line measurements of power production as well as HIRLAM predictions as input thus taking advantage of the auto-correlation, which is present in the power production for shorter pediction horizons. Statistical models are used to discribe the relationship between observed energy...

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

    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

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

    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.

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

    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

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

    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...

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

    Janstrup, Kira Hyldekær

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

  13. Confidence scores for prediction models

    Gerds, Thomas Alexander; van de Wiel, MA

    2011-01-01

    distinguish rival prediction models with similar prediction performances. Furthermore, on the subject level a confidence score may provide useful supplementary information for new patients who want to base a medical decision on predicted risk. The ideas are illustrated and discussed using data from cancer...... modelling strategy is applied to different training sets. For each modelling strategy we estimate a confidence score based on the same repeated bootstraps. A new decomposition of the expected Brier score is obtained, as well as the estimates of population average confidence scores. The latter can be used to...

  14. Modelling, controlling, predicting blackouts

    Wang, Chengwei; Baptista, Murilo S

    2016-01-01

    The electric power system is one of the cornerstones of modern society. One of its most serious malfunctions is the blackout, a catastrophic event that may disrupt a substantial portion of the system, playing havoc to human life and causing great economic losses. Thus, understanding the mechanisms leading to blackouts and creating a reliable and resilient power grid has been a major issue, attracting the attention of scientists, engineers and stakeholders. In this paper, we study the blackout problem in power grids by considering a practical phase-oscillator model. This model allows one to simultaneously consider different types of power sources (e.g., traditional AC power plants and renewable power sources connected by DC/AC inverters) and different types of loads (e.g., consumers connected to distribution networks and consumers directly connected to power plants). We propose two new control strategies based on our model, one for traditional power grids, and another one for smart grids. The control strategie...

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

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

  16. Melanoma Risk Prediction Models

    Developing statistical models that estimate the probability of developing melanoma cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  17. Hybrid Model for Early Onset Prediction of Driver Fatigue with Observable Cues

    Mingheng Zhang; Gang Longhui; Zhe Wang; Xiaoming Xu; Baozhen Yao; Liping Zhou

    2014-01-01

    This paper presents a hybrid model for early onset prediction of driver fatigue, which is the major reason of severe traffic accidents. The proposed method divides the prediction problem into three stages, that is, SVM-based model for predicting the early onset driver fatigue state, GA-based model for optimizing the parameters in the SVM, and PCA-based model for reducing the dimensionality of the complex features datasets. The model and algorithm are illustrated with driving experiment data a...

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

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

    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

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

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

  1. Predictive Models and Computational Embryology

    EPA’s ‘virtual embryo’ project is building an integrative systems biology framework for predictive models of developmental toxicity. One schema involves a knowledge-driven adverse outcome pathway (AOP) framework utilizing information from public databases, standardized ontologies...

  2. Babcock and Wilcox revisions to CONTEMPT, computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The analytical model used for the program is described. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc. 11 references. (U.S.)

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

    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

  4. Atmospheric modeling of radioactive material dispersion and health risk in Fukushima Daiichi nuclear power plants accident

    Highlights: ► The radioactive concentrations are treated as dynamical values. ► A possible nuclear accident is simulated for the prediction of atmospheric contaminations. ► The dangerous situations caused by radioisotope release could be announced to the public. ► In the future studies, some other variables are can be considered. - Abstract: The radioactive material dispersion is investigated in terms of the radioactive concentrations. The risk of the radioactive hazard material is important with respect to the public health. The prevailing westerlies region is modeled for the dynamical consequences, whereby the Fukushima nuclear disaster in Japan is modeled. The multiplications effects of the wind values and plume concentrations are obtained. Monte Carlo calculations are performed for wind speed and direction. In Seoul and Pusan, Korea, the Cs-137 has the highest value among the chemical radioactive materials Cs-137, I-131, and Sr-90. The time for highest concentration is shown to be around 48th hour in Seoul and 12th hour in Pusan. Cesium has the highest value in both cities, and iodine has the lowest value in both cities. The wind is assumed to determine the direction of movement. Therefore, the real values are believed to be lower than the calculated results. This modeling could be used for other industrial accident cases in chemical plants

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

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

    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.

  7. China's coal mine accident statistics analysis and one million tons mortality prediction

    Qiao Tong

    2016-01-01

    In order to study the general rule of coal mine accidents in China in recent years, the data of coal mine accident in 2011-2015 is analyzed. The mathematical statistics method is used to analyze the occurrence year, type, season and area of the accident. The results of analysis shows that the coal mine accident has been reduced year by year, and the frequency of gas explosion is the highest. The frequency and the number of deaths in the second quarter of the year are the highest; Guizhou p...

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

    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.

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

    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.

  10. Prediction of the onset of flow instability in the ETRR-2 research reactor under loss of flow accident

    In the present work, prediction of the onset of flow instability (OFI) in Egypt's second research reactor (ETRR-2) under loss of flow accident (LOFA) conditions due to loss of off-site power has been performed based on the model developed in previous work. Calculations are performed for LOFA with Scram due to low flow signal at 90% of the nominal flow where the time length that covers this transient is 4.8 seconds. Both, the best-estimate and conservative calculations are performed under the specified operating conditions and geometrical data of ETRR-2 for both, exponential and ramp pressure gradient change. The OFI locus is predicted and plotted against the flow velocity, exit coolant temperature and bubble detachment parameter for several heat fluxes. The results are analyzed and assessed in order to obtain the safety margins for OFI phenomenon that measure how far the operating conditions are from the OFI locus. It is found that, the safety margins for OFI phenomenon in the best-estimate calculation are 2.62 and 2.35 for steady state and LOFA transient respectively, while its values in the conservative calculation are 1.68 and 1.47, respectively. (orig.)

  11. Researching Effects of Drivers Features on Traffic Accidents: Kocaeli Model

    UÇKUN, Ceylan Gazi; ÇELİKKOL, Ethem Soner; TEKİN, Vasfı Nadir; ÇELİKKOL, Şimal

    2013-01-01

    In addition to environmental conditions, weather conditions and density, situations related to drivers are more effective on traffic accidents, according to available data. Regarding occurrence of traffic accidents, it is observed that point of view of drivers towards traffic rules and drivers’ compliance with these rules is not parallel. It is important to research the reasons that cause this situation. A normal person’s mental state does not change without any reason at traffic. It is clear...

  12. RESEARCHING EFFECTS OF DRIVERS FEATURES ON TRAFFIC ACCIDENTS: KOCAELİ MODEL

    CEYLAN GAZI UÇKUN; ETHEM SONER ÇELİKKOL; VASFI NADIR TEKİN; ŞIMAL ÇELİKKOL

    2013-01-01

    In addition to environmental conditions, weather conditions and density, situations related to drivers are more effective on traffic accidents, according to available data.Regarding occurrence of traffic accidents, it is observed that point of view of drivers towards traffic rules and drivers’ compliance with these rules is not parallel. It is important to research the reasons that cause this situation. A normal person’s mental state does not change without any reason at traffic. It is clear ...

  13. Evaluating the reliability of predictions made using environmental transfer models

    The development and application of mathematical models for predicting the consequences of releases of radionuclides into the environment from normal operations in the nuclear fuel cycle and in hypothetical accident conditions has increased dramatically in the last two decades. This Safety Practice publication has been prepared to provide guidance on the available methods for evaluating the reliability of environmental transfer model predictions. It provides a practical introduction of the subject and a particular emphasis has been given to worked examples in the text. It is intended to supplement existing IAEA publications on environmental assessment methodology. 60 refs, 17 figs, 12 tabs

  14. Predictive Modelling of Cellular Load

    Carolan, Emmett; McLoone, Seamus; Farrell, Ronan

    2015-01-01

    This work examines the temporal dynamics of cellular load in four Irish regions. Large scale underutilisation of network resources is identified both at the regional level and at the level of individual cells. Cellular load is modeled and prediction intervals are generated. These prediction intervals are used to put an upper bound on usage in a particular cell at a particular time. Opportunities for improvements in network utilization by incorporating these upper bounds on usage are identifie...

  15. Modelling of the hydrogen production during the reflooding phase in case of severe accident in a nuclear power plant reactor

    In 1979, the Three Mile Island (TMI) accident accelerated research activities in the field of severe accidents, i.e. accidents leading to a significant core degradation. Among the different computer codes developed in this scope, one of them is a scenario code, called Modular Accident Analysis Program (MAAP). It has been developed in the US and has been used by Electricite de France since 1991 to carry out safety analyses. In this thesis, only severe accidents that lead the core of a Pressurized Water Reactor to be partially or totally uncovered are considered. To avoid that such accidents get worse and lead to a radioactivity release into the environment, procedures imply massive water injections to flood the core. Different comparative studies showed that current computer codes, including the MAAP code, could not model correctly this phenomenon and, in particular, could not predict with accuracy the generation of hydrogen observed in experiments. In a certain range of concentrations, hydrogen and oxygen could recombine in an explosive manner. To prevent this risk in France, it has been decided to build passive auto-catalytic recombiners in the reactor containment building. Their design is strongly dependant on the hydrogen generation kinetics that is estimated with such computer codes. This thesis aims at gathering the state-of-the-art knowledge from a literature review, analysing current models in the MAAP4 code, developing new models and validating them against data from the TMI accident and from the QUENCH experiments (carried out in Forschungszentrum Karlsruhe, Germany). The main results of this research led us to change the oxidation correlations that apply at high temperature in the MAAP4 code and to add two new models. The first one is a simplified two-phase flow thermal-hydraulics model that improves the calculation of the cladding temperature axial profile; the second model takes into consideration the increase of the surface likely to get oxidized after

  16. A contrail cirrus prediction model

    U. Schumann

    2012-05-01

    Full Text Available A new model to simulate and predict the properties of a large ensemble of contrails as a function of given air traffic and meteorology is described. The model is designed for approximate prediction of contrail cirrus cover and analysis of contrail climate impact, e.g. within aviation system optimization processes. The model simulates the full contrail life-cycle. Contrail segments form between waypoints of individual aircraft tracks in sufficiently cold and humid air masses. The initial contrail properties depend on the aircraft. The advection and evolution of the contrails is followed with a Lagrangian Gaussian plume model. Mixing and bulk cloud processes are treated quasi analytically or with an effective numerical scheme. Contrails disappear when the bulk ice content is sublimating or precipitating. The model has been implemented in a "Contrail Cirrus Prediction Tool" (CoCiP. This paper describes the model assumptions, the equations for individual contrails, and the analysis-method for contrail-cirrus cover derived from the optical depth of the ensemble of contrails and background cirrus. The model has been applied for a case study and compared to the results of other models and in-situ contrail measurements. The simple model reproduces a considerable part of observed contrail properties. Mid-aged contrails provide the largest contributions to the product of optical depth and contrail width, important for climate impact.

  17. A contrail cirrus prediction model

    U. Schumann

    2011-11-01

    Full Text Available A new model to simulate and predict the properties of a large ensemble of contrails as a function of given air traffic and meteorology is described. The model is designed for approximate prediction of contrail cirrus cover and analysis of contrail climate impact, e.g. within aviation system optimization processes. The model simulates the full contrail life-cycle. Contrail segments form between waypoints of individual aircraft tracks in sufficiently cold and humid air masses. The initial contrail properties depend on the aircraft. The advection and evolution of the contrails is followed with a Lagrangian Gaussian plume model. Mixing and bulk cloud processes are treated quasi analytically or with an effective numerical scheme. Contrails disappear when the bulk ice content is sublimating or precipitating. The model has been implemented in a "Contrail Cirrus Prediction Tool" (CoCiP. This paper describes the model assumptions, the equations for individual contrails, and the analysis-method for contrail-cirrus cover derived from the optical depth of the ensemble of contrails and background cirrus. The model has been applied for a case study and compared to the results of other models and in-situ contrail measurements. The simple model reproduces a considerable part of observed contrail properties. Mid-aged contrails provide the largest contributions to the product of optical depth and contrail width, important for climate impact.

  18. Fuel thermal/mechanical behaviour under loss of coolant accident conditions as predicted by the FACTAR code

    FACTAR (Fuel And Channel Temperature And Response) is a computer code developed to simulate the transient thennal and mechanical behaviour of 37-element or 28-element fuel bundles within a single CANDU fuel channel for moderate (ie., sheath temperatures less than the melting point of Zircaloy) loss-of-coolant accident (LOCA) conditions including transition and large break LOCAs with emergency coolant injection assumed available. FACTAR's predictions of fuel temperature and sheath failure times are used for subsequent assessment of fission product releases and fuel string expansion. In this paper, model capabilities and calculated quantities of the code are summarised. The results from overly severe test cases are presented in order to clearly demonstrate the effect on calculated fuel channel behaviour of a mechanistic assessment of fuel-to-sheath heat transfer, and the impact of using a diffusion-limited model for Zircaloy/steam reaction (i.e., FROM) as opposed to a reaction rate correlation, coupled with the assumption of unlimited steam supply. (author)

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

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

    2012-04-01

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

  20. CFD modeling of debris melting phenomena during late phase Candu 6 severe accident

    The objective of this paper was to study the phase change of the debris formed on the Candu 6 calandria bottom in a postulated accident sequence. The molten pool and crust formation were studied employing the Ansys-Fluent code. The 3D model using Large Eddy Simulation (LES) predicts the conjugate, radiative and convective heat transfer inside and from the corium pool. LES simulations require a very fine grid to capture the crust formation and the free convection flow. This aspect (fine mesh requirement) correlated with the long transient has imposed the use of a slice from the 3D calandria geometry in order not to exceed the computing resources. The preliminary results include heat transfer coefficients, temperature profiles and heat fluxes through calandria wall. From the safety point of view it is very important to maintain a heat flux through the wall below the CHF assuring the integrity of the calandria vessel. This can be achieved by proper cooling of the tank water which contains the vessel. Also, transient duration can be estimated being important in developing guidelines for severe accidents management. The debris physical structure and material properties have large uncertainties in the temperature range of interest. Thus, further sensitivity studies should be carried out in order to better understand the influence of these parameters on this complex phenomenon. (authors)

  1. A new modelling approach for containment event tree construction -Accident progression stage event tree method

    The Accident Progression Stage Event Tree (APSET) method presented here is a new modelling approach for construction of comprehensive and concise containment event trees to describe physical processes inside containment and accident mitigation actions, yet provide enough detail to analyze important factors for containment responses to severe accidents. In this approach, the accident progression is generally divided into four accident stages, i.e., Pre-stage for Core-melt, Core-melt Progression Stage, Debris Exit Stage, and Long-term Progression Stage, to reflect the timing of containment failure. Physical phenomena which challenge the containment integrity and accident mitigation actions are chronologically represented in event trees for each stage. Event trees for two successive stages are cross-linked by interface parameter. The interface parameter is defined as a set of plant conditions that have a significant influence on physical processes in the subsequent stage. By quantifying the containment event trees constructed with the APSET method, the respective conditional probabilities of the containment failure modes and the accident termination can be calculated stage by stage for each core melt accident sequence. The quantification results provide the characteristics of each core melt sequence on containment responses such as a dominant containment failure mode, its timing, and the effectiveness of mitigation actions. The usefulness of the APSET method was demonstrated through its application to a containment event tree analysis for BWR with MARK-II containment. (author). 11 refs., 2 tabs., 4 figs

  2. A method for modeling and analysis of directed weighted accident causation network (DWACN)

    Zhou, Jin; Xu, Weixiang; Guo, Xin; Ding, Jing

    2015-11-01

    Using complex network theory to analyze accidents is effective to understand the causes of accidents in complex systems. In this paper, a novel method is proposed to establish directed weighted accident causation network (DWACN) for the Rail Accident Investigation Branch (RAIB) in the UK, which is based on complex network and using event chains of accidents. DWACN is composed of 109 nodes which denote causal factors and 260 directed weighted edges which represent complex interrelationships among factors. The statistical properties of directed weighted complex network are applied to reveal the critical factors, the key event chains and the important classes in DWACN. Analysis results demonstrate that DWACN has characteristics of small-world networks with short average path length and high weighted clustering coefficient, and display the properties of scale-free networks captured by that the cumulative degree distribution follows an exponential function. This modeling and analysis method can assist us to discover the latent rules of accidents and feature of faults propagation to reduce accidents. This paper is further development on the research of accident analysis methods using complex network.

  3. Predicting Abraham model solvent coefficients

    Bradley, Jean-Claude; Abraham, Michael H; Acree, William E; Lang, Andrew SID

    2015-01-01

    Background The Abraham general solvation model can be used in a broad set of scenarios involving partitioning and solubility, yet is limited to a set of solvents with measured Abraham coefficients. Here we extend the range of applicability of Abraham’s model by creating open models that can be used to predict the solvent coefficients for all organic solvents. Results We created open random forest models for the solvent coefficients e, s, a, b, and v that had out-of-bag R2 values of 0.31, 0.77...

  4. Modeling the early-phase redistribution of radiocesium fallouts in an evergreen coniferous forest after Chernobyl and Fukushima accidents

    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

  5. Modeling the early-phase redistribution of radiocesium fallouts in an evergreen coniferous forest after Chernobyl and Fukushima accidents

    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

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

    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)

  7. Preconditioned Continuation Model Predictive Control

    Knyazev, Andrew; Fujii, Yuta; Malyshev, Alexander,

    2015-01-01

    Model predictive control (MPC) anticipates future events to take appropriate control actions. Nonlinear MPC (NMPC) describes systems with nonlinear models and/or constraints. A Continuation/GMRES Method for NMPC, suggested by T. Ohtsuka in 2004, uses the GMRES iterative algorithm to solve a forward difference approximation $Ax=b$ of the Continuation NMPC (CNMPC) equations on every time step. The coefficient matrix $A$ of the linear system is often ill-conditioned, resulting in poor GMRES conv...

  8. The grey interrelation analysis and trend prediction on the safety accident in Kailun Coal Mine

    Yang, Z.; Ding, Y.; Zhao, C. [Kailun (Group) Limited Liability Corporation, Tangshan (China)

    2003-02-01

    The man-machine-environment systems in Kailuan Coal Mines is taken as the object of study to make the grey interrelation analysis for coal mine accidents and related factors by integrating the Grey System Theory with actual coal mine production. It also forecasts the accident development trend in coalmine in accordance with the accident statistics of coalmine by means of the grey forecast method. The injury rate per 1000 persons in Jinggezhuang Coal Mine in 2001 and 2002 was forecast and the results were 8.1043 and 7.7033 respectively. The process and the result in the analysis and forecast indicate that the method is simple and easy to use, and the result is reliable. The method and result of the study provide the theoretical reference for the quantitative study in coalmine accidents, as well as the basis for decision-making on safety management of coal enterprise. 3 refs., 4 tabs.

  9. Coping and health status predicts PTSD 12 months after a serious motor vehicle accident

    Pires, Tânia Sofia Fernandes; Maia, Ângela

    2011-01-01

    Background: Maladjusted coping strategies after motor vehicle accidents (MVA) can contribute to the development of psychological symptoms, as PTSD. Methods: Measures of Acute Stress Disorder, PTSD scale, Coping, Social Support and physical health were used to evaluate 101MVA victims with serious injuries 5 days, 4 and 12 months after the accident Findings: 67% of the participants had ASD (T1), 58% had PTSD at T2 and 47% had PTSD at T3. Victims that report more general coping strategies...

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

    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. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code

  12. Base irradiation simulation and its effect on fuel behavior prediction by TRANSURANUS code: Application to reactivity initiated accident condition

    Lisovyy, Oleksandr, E-mail: o.lisovyy@dimnp.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); Cherubini, Marco, E-mail: m.cherubini@ing.unipi.it [NINE, Via Livornese 1291, Pisa 56122 (Italy); Lazzerini, Davide, E-mail: d.lazzerini@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy); D’Auria, Francesco, E-mail: f.dauria@ing.unipi.it [GRNSPG-UNIPI, Via Livornese 1291, Pisa 56122 (Italy)

    2015-03-15

    Highlights: • Selection of parameters for analysis. • Base irradiation simulation of rods fabricated by ENUSA. • Boundary condition implementation using restart options. • RIA simulation of CABRI test CIP3-1. • Sensitivity analysis performance. - Abstract: The purpose of the present paper is to investigate the impact of the base irradiation simulation for predicting fuel behavior under Reactivity Initiated Accident (RIA) conditions. A RIA is a scenario challenging the fuel integrity and consequently, devoted experimental campaigns and related code simulations have been extensively performed. In all experiments in which irradiated fuel is tested, the experiment is preceded by in reactor period, i.e. the base irradiation. In the present paper the considered RIA experiment is CIP3-1 performed in CABRI reactor (part of the OECD/NEA WGFS benchmark); a discussion about the relevance of the base irradiation simulation is presented. Such a work is conducted by sensitivities calculation in which a single parameter, among a preselected set, is changed. The range of variation of such parameters is either supplied within the selected RIA test specification or is taken from typical values available in the open literature. All mentioned calculations have been performed developing a specific model in TRANSURANUS code.

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

    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.

  14. Radioecological zoning of territory and model of territory for monitoring of agrosphere after heavy accident at the NPP

    To improve the effectiveness of responses to severe accident in the field of population and agricultural production before the accident, proposed to prevent collect and analyze cartographic, statistical, environmental and others. The information needed to predict and assess the radiological situation. The methodology of radio-ecological zoning of the territory contaminated with radioactive fallout, using GIS technology, which was based on landscape-basin principle. A model of the territory, taxonomic units which are elements of the landscape or objects of agricultural land use. The river pond is a primary objective of the existing structural unit of the territory. The main characteristics are the type of soil, the type of terrain and the type of underlying surface. The application model provides the coordination of spatial and temporal distribution of characteristics, coupled models of atmospheric diffusion and migration of radionuclides on the chain ''soil - plants - animals - Products - man'' and dosimetric models to determine countermeasures that may be necessary after the accident. To forecast the radiation environment on the track used by the accidental release of the authors developed a model of atmospheric transport of radionuclides, aeral and root of plant contamination

  15. Application of Westinghouse NEXUS/ANC9 cross-section model for PWR accident analyses

    NEXUS/ANC9 is the latest licensed PWR core design code system developed by Westinghouse. This system has demonstrated capabilities of modeling advanced core designs with improved accuracy in core reactivity and power distribution predictions. NEXUS/ANC9 system is being rolled out to replace the current APA system (ALPHA/PHOENIX-P/ANC) for routine core calculations. In addition to the standard core design calculations, investigations are underway to explore the possibility to expand the NEXUS/ANC9 application for safety analysis, especially at accident conditions. The main focus of the investigation is the evaluation of the NEXUS/ANC9 cross-section representation model conditions like high void and significant change of core pressure. Comparisons of the predicted parameters among ANC9, PARAGON lattice code and MCNP calculations are presented. The results show that NEXUS/ANC9 is able to model the cross-section behavior and accurately reproduce lattice code results at all simulated conditions. (author)

  16. Failure prediction model: Model napovedovanja odpovedi:

    Čelan, Štefan; Težak, Oto; Žižek, Adolf

    2002-01-01

    Preventative maintenance is vital for delicate technical products. Electronic components or the whole system must be changed, and thus need a good model that will indicate failure accurately. In this paper a stochastic stress-strength quantitative model is presented, folowing the five original hypothesis. Proposed new model of failure prediction could be used by the system maintenance. Failure risk could be instantaneosly calculated. The given theory considers the influences of stress on the ...

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

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

    2013-11-15

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

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

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

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

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

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

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

  1. A model for the analysis of loss of decay heat removal accident in MTR pool type research reactors

    During a loss of coolant accident leading to total emptying of the reactor pool, the decay heat could be removed through air natural convection. However, under partial pool emptying the core is partially submerged and the coolant circulation inside the fuel element could no more be possible. In such conditions, a core overheat take place, and the heat is essentially diffused from the core to its periphery by combined thermal radiation and conduction. In order to predict fuel element temperature evolution under such conditions a mathematical model is performed. The model is based on a three dimensional geometry and takes into account a variety of core configurations including fuel elements (standard and control), reflector elements and grid plates. The homogeneous flow model is used and the time and space dependent non-linear partial differential fluid conservation equations are solved using a semi-implicit finite difference method. Preliminary tests of the developed model were made by considering a series of hypothetical accidents. In the current framework a loss of decay heat removal accidents in the IAEA benchmark open pool MTR-type research reactor is considered. It is shown that in the case of a low core immersion height no water boiling is observed and the fuel surface temperature rise remains below the melting point of the aluminium cladding. (author)

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

    , a graph representing the time to failure as a function of the pressure level and the heat flux intensity has been determined; such information will be used in our probabilistic safety assessment and severe accident management analyses. Another motivation for the development of simplified models in IRSN is to obtain a simplified but well-predicting code that can then be integrated into integral severe accident computer codes. (authors)

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

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

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

    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...

  5. Transportation accidents

    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

  6. Radiation effects on the population of Belarus after the Chernobyl accident and the prediction of stochastic effects

    Evaluation of conditions of exposure during the post-accident period makes it possible to identify two periods in the radiation exposure of Belarus's population. As a result of our investigations we obtained data about doses for four different categories in the exposed population: people who lived in the contaminated territories without evacuation and relocation; evacuated people; cleanup workers (''liquidators''); and people who were exposed in childhood, especially for thyroid exposure. The total doses for these categories in different time periods were analyzed. Evaluation of doses received by the Belarusian population due to the Chernobyl accident shows no evidence of doses, that could lead to the deterministic consequences of radiation exposure. For all exposed groups we made predictions about different types of stochastic consequences of exposure. 10 refs, 2 tabs

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

    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.

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

    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

  9. Hybrid Model for Early Onset Prediction of Driver Fatigue with Observable Cues

    Mingheng Zhang

    2014-01-01

    Full Text Available This paper presents a hybrid model for early onset prediction of driver fatigue, which is the major reason of severe traffic accidents. The proposed method divides the prediction problem into three stages, that is, SVM-based model for predicting the early onset driver fatigue state, GA-based model for optimizing the parameters in the SVM, and PCA-based model for reducing the dimensionality of the complex features datasets. The model and algorithm are illustrated with driving experiment data and comparison results also show that the hybrid method can generally provide a better performance for driver fatigue state prediction.

  10. Monitoring severe accidents using AI techniques

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

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

    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.)

  12. A dynamical model predicting the transport of I-131 through the deposition pasture cow milk pathway

    A dynamical model predicting the transport of I-131 through the atmospherical deposition-pasture-cow-milk pathway has been developed and validated using data collected in a specific site (a little farm in Anguillara - Rome) during the Chernobyl accident. The main factor affecting the uncertainty of the results of the model are discussed

  13. Severe accident modeling and offsite dose consequence evaluations for nuclear power plant emergency planning

    We have investigated the roles of Firewater Addition System and Passive Flooder in ABWR severe accidents, such as LOCA and SBO. The results are apparent that Firewater System is vital in the highly unlikely situation where all AC are lost. Also in this paper, we present EPZDose, an effective and faster-than-real time code for offsite dose consequences predictions and evaluations. Illustrations with the release from our severe accident scenario show friendly and informative user's interface for supporting decision makings in nuclear emergency situations. (author)

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

    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.

  15. Pin-by-pin modeling of fuel cycle and reactivity initiated accidents in LWR

    This study deals with validation results for pin-by-pin methods to model fuel cycle and reactivity initiated accidents (RIAs) in LWR. Both methods are based on a heterogeneous pin-by-pin reactor model, realized in the BARS code. Validation results are presented for separate steps of WWER fuel cycle modeling. Features and advantages of a pin-by-pin approach for modeling of LWR RIA shown on the basis of calculations of control rod ejection accidents (REAs) in South Ukrainian NPP Unit 1 WWER-1000 and Three Mile Island Unit 1 (TMI-1) PWR at the end of cycles. Calculations were performed using the coupled RELAP-BARS code. Effects of pin-by-pin power and burnup distribution on estimation of the accident consequences are considered. (Authors)

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

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

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

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

    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.)

  19. Guide-lines for an early evaluation of a nuclear accident, calculated with the computer model park

    For a nuclear accident where large areas are contaminated, it is necessary to predict the exposure of the population as early as possible in order to plan appropriate countermeasures. The radioecological computer model PARK (Program System for the Assessment and Mitigation of Radiological Consequences) is part of the German decision support system IMIS (Integrated Measurement- and Information System for the Surveillance of Environmental Radioactivity) for a fast assessment of contaminations and doses. In this paper PARK is used to investigate the dose relevance of the exposure pathways, of ingested radionuclides, and of foodstuffs in relation to the date of the event. (author)

  20. A cladding failure model for fuel rods subjected to operational and accident transients

    Concerns about high burnup effects on cladding integrity during operational and accident transients have been invoked by licensing authorities in the United States of America, Europe and Japan as potentially limiting for burnup extension. Transient experiments recently conducted in France and Japan to simulate reactivity initiation accidents (RIAs) in light water reactors have shown that high burnup fuel rods can fail at enthalpy levels well below the current licensing limits. Analytical research conducted by EPRI during the last few years, in support of the RIA tests evaluation, has led to the development of a cladding failure model for reactor transients, including RIA and power oscillation events in boiling water reactors known as ATWS (anticipated transient without scram). The model is incorporated in EPRI's fuel behavior code FALCON, which is the modern version of the FREY code that was presented in previous IAEA fuel behavior meeting. The most distinguishing feature of the model is that it computes the mechanical energy locally at material points in the cladding as function of time during the transient event, from which the failure location and failure time are predicted. The database for the model consists of stress-strain data obtained from mechanical property tests for cladding tubes as function of fast fluence, temperature, hydrogen concentration and material type. From this data, the material's capacity, or resistance to failure, is formulated as the total (elastic+plastic) mechanical energy per unit volume that can be absorbed by the cladding before it can fail, and is termed the critical strain energy density (CSED). The FALCON code calculates the strain energy density (SED) that a transient event can deliver to the cladding through PCMI and internal pressure loading, which is then compared to the CSED for failure determination. Clearly, the complete stress and strain states enter into the calculation of the SED, and therefore, all three true

  1. Prediction of mass fraction of agglomerated debris in a LWR severe accident

    Ex-vessel termination of accident progression in Swedish type Boiling Water Reactors (BWRs) is contingent upon efficacy of melt fragmentation and solidification in a deep pool of water below reactor vessel. When liquid melt reaches the bottom of the pool it can create agglomerated debris and “cake” regions that increase hydraulic resistance of the bed and affect coolability of the bed. This paper discusses development and application of a conservative-mechanistic approach to quantify mass fractions of agglomerated debris. Experimental data from the DEFOR-A (Debris Bed Formation and Agglomeration) tests with high superheat of binary oxidic simulant material melt is used for validation of the methods. Application of the approach to plant accident analysis suggests that melt superheat has less significant influence on agglomeration of the debris than jet penetration depth. The paper also discusses the impact of the uncertainty in the jet disintegration and penetration behavior on the agglomeration mode map. (author)

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

    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.

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

    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

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

    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

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

    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

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

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

  7. Advanced evacuation model managed through fuzzy logic during an accident in LNG terminal

    Evacuation of people located inside the enclosed area of an LNG terminal is a complex problem, especially considering that accidents involving LNG are potentially very hazardous. In order to create an evacuation model managed through fuzzy logic, extensive influence must be generated from safety analyses. A very important moment in the optimal functioning of an evacuation model is the creation of a database which incorporates all input indicators. The output result is the creation of a safety evacuation route which is active at the moment of the accident. (Author)

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

    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

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

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

  10. DC Motor Control Predictive Models

    Ravinesh Singh

    2006-01-01

    Full Text Available DC motor speed and position controls are fundamental in vehicles in general and robotics in particular. This study presents a mathematical model for correlating the interactions of some DC motor control parameters such as duty cycle, terminal voltage, frequency and load on some responses such as output current, voltage and speed by means of response surface methodology. For this exercise, a five-level full factorial design was chosen for experimentation using a peripheral interface controller (PIC-based universal pulse width modulation (PWM H-Bridge motor controller built in-house. The significance of the mathematical model developed was ascertained using regression analysis method. The results obtained show that the mathematical models are useful not only for predicting optimum DC motor parameters for achieving the desired quality but for speed and position optimization. Using the optimal combination of these parameters is useful in minimizing the power consumption and realization of the optimal speed and invariably position control of DC motor operations.

  11. Predictive Modeling of Tokamak Configurations*

    Casper, T. A.; Lodestro, L. L.; Pearlstein, L. D.; Bulmer, R. H.; Jong, R. A.; Kaiser, T. B.; Moller, J. M.

    2001-10-01

    The Corsica code provides comprehensive toroidal plasma simulation and design capabilities with current applications [1] to tokamak, reversed field pinch (RFP) and spheromak configurations. It calculates fixed and free boundary equilibria coupled to Ohm's law, sources, transport models and MHD stability modules. We are exploring operations scenarios for both the DIII-D and KSTAR tokamaks. We will present simulations of the effects of electron cyclotron heating (ECH) and current drive (ECCD) relevant to the Quiescent Double Barrier (QDB) regime on DIII-D exploring long pulse operation issues. KSTAR simulations using ECH/ECCD in negative central shear configurations explore evolution to steady state while shape evolution studies during current ramp up using a hyper-resistivity model investigate startup scenarios and limitations. Studies of high bootstrap fraction operation stimulated by recent ECH/ECCD experiments on DIIID will also be presented. [1] Pearlstein, L.D., et al, Predictive Modeling of Axisymmetric Toroidal Configurations, 28th EPS Conference on Controlled Fusion and Plasma Physics, Madeira, Portugal, June 18-22, 2001. * Work performed under the auspices of the U.S. Department of Energy by the University of California, Lawrence Livermore National Laboratory under contract No. W-7405-Eng-48.

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

    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

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

    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

  14. Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

    NONE

    2004-07-01

    This file is devoted to the Chernobyl accident. It is divided in four parts. The first part concerns the accident itself and its technical management. The second part is relative to the radiation doses and the different contaminations. The third part reports the sanitary effects, the determinists ones and the stochastic ones. The fourth and last part relates the consequences for the other European countries with the case of France. Through the different parts a point is tackled with the measures taken after the accident by the other countries to manage an accident, the cooperation between the different countries and the groups of research and studies about the reactors safety, and also with the international medical cooperation, specially for the children, everything in relation with the Chernobyl accident. (N.C.)

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

    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

  16. Containment severe accident thermohydraulic phenomena

    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)

  17. Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

    Nalbandyan, A.; Ytre-Eide, M.A.; Thoerring, H.; Liland, A.; Bartnicki, J.; Balonov, M.

    2012-06-15

    The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

  18. Potential consequences in Norway after a hypothetical accident at Leningrad nuclear power plant. Potential release, fallout and predicted impacts on the environment

    The report describes different hypothetical accident scenarios at the Leningrad nuclear power plant for both RBMK and VVER-1200 reactors. The estimated release is combined with different meteorological scenarios to predict possible fallout of radioactive substances in Norway. For a hypothetical catastrophic accident at an RBMK reactor combined with a meteorological worst case scenario, the consequences in Norway could be considerable. Foodstuffs in many regions would be contaminated above the food intervention levels for radioactive cesium in Norway. (Author)

  19. Severe Nuclear Accident Program (SNAP) - a real time model for accidental releases

    The model: Several Nuclear Accident Program (SNAP) has been developed at the Norwegian Meteorological Institute (DNMI) in Oslo to provide decision makers and Government officials with real-time tool for simulating large accidental releases of radioactivity from nuclear power plants or other sources. SNAP is developed in the Lagrangian framework in which atmospheric transport of radioactive pollutants is simulated by emitting a large number of particles from the source. The main advantage of the Lagrangian approach is a possibility of precise parameterization of advection processes, especially close to the source. SNAP can be used to predict the transport and deposition of a radioactive cloud in e future (up to 48 hours, in the present version) or to analyze the behavior of the cloud in the past. It is also possible to run the model in the mixed mode (partly analysis and partly forecast). In the routine run we assume unit (1 g s-1) emission in each of three classes. This assumption is very convenient for the main user of the model output in case of emergency: Norwegian Radiation Protection Agency. Due to linearity of the model equations, user can test different emission scenarios as a post processing task by assigning different weights to concentration and deposition fields corresponding to each of three emission classes. SNAP is fully operational and can be run by the meteorologist on duty at any time. The output from SNAP has two forms: First on the maps of Europe, or selected parts of Europe, individual particles are shown during the simulation period. Second, immediately after the simulation, concentration/deposition fields can be shown every three hours of the simulation period as isoline maps for each emission class. In addition, concentration and deposition maps, as well as some meteorological data, are stored on a public accessible disk for further processing by the model users

  20. Modelling of fission product release behavior from HTR spherical fuel elements under accident conditions

    Computer codes for modelling the fission product release behavior of spherical fuel elements for High Temperature Reactors (HTR) have been developed for the purpose of being used in risk analyses for HTRs. An important part of the validation and verification procedure for these calculation models is the theoretical investigation of accident simulation experiments which have been conducted in the KueFA test facility in the Hot Cells at KFA. The paper gives a presentation of the basic modeling and the calculational results of fission product release from modern German HTR fuel elements in the temperature range 1600-1800 deg. C using the TRISO coated particle failure model PANAMA and the diffusion model FRESCO. Measurements of the transient release behavior for cesium and strontium and of their concentration profiles after heating have provided informations about diffusion data in the important retention barriers of the fuel: silicon carbide and matrix graphite. It could be shown that the diffusion coefficients of both cesium and strontium in silicon carbide can significantly be reduced using a factor in the range of 0.02 - 0.15 compared to older HTR fuel. Also in the development of fuel element graphite, a tendency towards lower diffusion coefficients for both nuclides can be derived. Special heating tests focussing on the fission gases and iodine release from the matrix contamination have been evaluated to derive corresponding effective diffusion data for iodine in fuel element graphite which are more realistic than the iodine transport data used so far. Finally, a prediction of krypton and cesium release from spherical fuel elements under heating conditions will be given for fuel elements which at present are irradiated in the FRJ2, Juelich, and which are intended to be heated at 1600/1800 deg. C in the KueFA furnace in near future. (author). 7 refs, 11 figs

  1. A dynamic model for evaluating radionuclide distribution in forests from nuclear accidents

    The Chernobyl Nuclear Power Plant accident in 1986 caused radionuclide contamination in most countries in Eastern and Western Europe. A prime example is Belarus where 23% of the total land area received chronic levels; about 1.5 X 106 ha of forested lands were contaminated with 40-190 kBq m-2 and 2.5 X 104 ha received greater than 1,480 kBq m-2 of 137Cs and other long-lived radionuclides such as 90Sr and 239,240Pu. Since the radiological dose to the forest ecosystem will tend to accumulate over long time periods (decades to centuries), we need to determine what countermeasures can be taken to limit this dose so that the affected regions can, once again, safely provide habitat and natural forest products. To address some of these problems, our initial objective is to formulate a generic model, FORESTPATH, which describes the major kinetic processes and pathways of radionuclide movement in forests and natural ecosystems and which can be used to predict future radionuclide concentrations. The model calculates the time-dependent radionuclide concentrations in different compartments of the forest ecosystem based on the information available on residence half-times in two forest types: coniferous and deciduous. The results show that the model reproduces well the radionuclide cycling pattern found in the literature for deciduous and coniferous forests. Variability analysis was used to access the relative importance of specific parameter values in the generic model performance. The FORESTPASTH model can be easily adjusted for site-specific applications. 92 refs., 5 figs., 6 tabs

  2. Development of super simulator `IMPACT`. Pt. 2. Modeling of severe accident phenomena and initial verification tests

    Miyagi, Kazumi; Vierow, K.M.; Naitoh, Masanori [Nuclear Power Engineering Corp., Tokyo (Japan); Hidaka, Masataka; Susuki, Akira; Ishida, Naoyuki; Yamagishi, Makoto; Abe, Nobuaki

    1998-05-01

    IMPACT employs advanced methods of physical modeling and numerical computation and can simulate a wide spectrum of scenarios ranging from normal operation to hypothetical, severe accidents. The simulator models major phenomena in the accident such as thermal hydraulics in the reactor cooling system (RCS), heat up of fuel rods, core melt, molten core relocation (freezing, slumping etc.), debris cooling in lower plenum, fission product (FP)s release and transport in RCS, steam explosion, debris/concrete interaction, thermal hydraulics in the containment vessel (CV), FPs transport in the CV. The modeling and combination of these phenomena could supply the ability to simulate the severe accident progress at Light Water Reactor. Initiated in fiscal year 1993, the project`s conceptual and detailed design phases have been completed and coding and verification phases are in progress. In the several analysis modules, verification studies for some modules are under way, and the steam explosion analysis module and the debris coolability analysis module are examined against typical experimental results to confirm the ability of each model. The premixing submodule for analysis of steam explosion phenomena under severe accident conditions has been completed, and was shown to simulate the MIXA tests well. Calculation results of debris spreading model in debris cooling process are compared with the experimental results and calculated average location of the spearhead at each time shows good agreement with the experimental observation, though spearhead shapes are different. (author)

  3. Development of super simulator 'IMPACT'. Pt. 2. Modeling of severe accident phenomena and initial verification tests

    IMPACT employs advanced methods of physical modeling and numerical computation and can simulate a wide spectrum of scenarios ranging from normal operation to hypothetical, severe accidents. The simulator models major phenomena in the accident such as thermal hydraulics in the reactor cooling system (RCS), heat up of fuel rods, core melt, molten core relocation (freezing, slumping etc.), debris cooling in lower plenum, fission product (FP)s release and transport in RCS, steam explosion, debris/concrete interaction, thermal hydraulics in the containment vessel (CV), FPs transport in the CV. The modeling and combination of these phenomena could supply the ability to simulate the severe accident progress at Light Water Reactor. Initiated in fiscal year 1993, the project's conceptual and detailed design phases have been completed and coding and verification phases are in progress. In the several analysis modules, verification studies for some modules are under way, and the steam explosion analysis module and the debris coolability analysis module are examined against typical experimental results to confirm the ability of each model. The premixing submodule for analysis of steam explosion phenomena under severe accident conditions has been completed, and was shown to simulate the MIXA tests well. Calculation results of debris spreading model in debris cooling process are compared with the experimental results and calculated average location of the spearhead at each time shows good agreement with the experimental observation, though spearhead shapes are different. (author)

  4. Real-time EEG-based detection of fatigue driving danger for accident prediction.

    Wang, Hong; Zhang, Chi; Shi, Tianwei; Wang, Fuwang; Ma, Shujun

    2015-03-01

    This paper proposes a real-time electroencephalogram (EEG)-based detection method of the potential danger during fatigue driving. To determine driver fatigue in real time, wavelet entropy with a sliding window and pulse coupled neural network (PCNN) were used to process the EEG signals in the visual area (the main information input route). To detect the fatigue danger, the neural mechanism of driver fatigue was analyzed. The functional brain networks were employed to track the fatigue impact on processing capacity of brain. The results show the overall functional connectivity of the subjects is weakened after long time driving tasks. The regularity is summarized as the fatigue convergence phenomenon. Based on the fatigue convergence phenomenon, we combined both the input and global synchronizations of brain together to calculate the residual amount of the information processing capacity of brain to obtain the dangerous points in real time. Finally, the danger detection system of the driver fatigue based on the neural mechanism was validated using accident EEG. The time distributions of the output danger points of the system have a good agreement with those of the real accident points. PMID:25541095

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

    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

  6. Instrument Fault Detection Sensitivity of an Empirical Model under Accident Condition in NPPs

    After the recent accident in Fukushima, Japan, it has been proven that we cannot obtain fully reliable information from instruments during severe accident conditions. Although the reactor core really melted down, the RV water level indicator showed a more optimistic value than the actual conditions. Accordingly, plant operators were under the misunderstanding that the core was not exposed. This caused confusion for the incident response. Therefore, it is necessary to be equipped with a function that informs operators of the status of the instrument integrity in real time. If plant operators verify that the instruments are working properly during accident conditions, they able to make safer decisions. In an effort to solve this problem, we considered an empirical model using a Process Equipment Monitoring (PEM) tool as a method of instrument diagnosis in a nuclear power plant

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

    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

  8. Risk factors associated with bus accident severity in the United States: A generalized ordered logit model

    Kaplan, Sigal; Prato, Carlo Giacomo

    2012-01-01

    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...

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

    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.)

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

    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.

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

    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

  12. A nuclear plant accident diagnosis method to support prediction of errors of commission

    The identification and mitigation of operator errors of commission (EOCs) continue to be a major focus of nuclear plant human reliability research. Current Human Reliability Analysis (HRA) methods for predicting EOCs generally rely on the availability of operating procedures or extensive use of expert judgment. Consequently, an analysis for EOCs cannot easily be performed for actions that may be taken outside the scope of the operating procedures. Additionally, current HRA techniques rarely capture an operator's 'creative' problem-solving behavior. However, a nuclear plant operator knowledge base developed for the use with the IDAC (Information, Decision, and Action in Crew context) cognitive model shows potential for addressing these limitations. This operator knowledge base currently includes an event-symptom diagnosis matrix for a pressurized water reactor (PWR) nuclear plant. The diagnosis matrix defines a probabilistic relationship between observed symptoms and plant events that models the operator's heuristic process for classifying a plant state. Observed symptoms are obtained from a dynamic thermal-hydraulic plant model and can be modified to account for the limitations of human perception and cognition. A fuzzy-logic inference technique is used to calculate the operator's confidence, or degree of belief, that a given plant event has occurred based on the observed symptoms. An event diagnosis can be categorized as either: (a) a generalized flow imbalance of basic thermal-hydraulic properties (e.g., a mass or energy flow imbalance in the reactor coolant system), or (b) a specific event type, such as a steam generator tube rupture or a reactor trip. When an operator is presented with incomplete or contradictory information, this diagnosis approach provides a means to identify situations where an operator might be misled to perform unsafe actions based on an incorrect diagnosis. This knowledge base model could also support identification of potential EOCs when

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

    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.

  14. A risk-based evaluation of the impact of key uncertainties on the prediction of severe accident source terms - STU

    The purpose of this project is to address the key uncertainties associated with a number of fission product release and transport phenomena in a wider context and to assess their relevance to key severe accident sequences. This project is a wide-based analysis involving eight reactor designs that are representative of the reactors currently operating in the European Union (EU). In total, 20 accident sequences covering a wide range of conditions have been chosen to provide the basis for sensitivity studies. The appraisal is achieved through a systematic risk-based framework developed within this project. Specifically, this is a quantitative interpretation of the sensitivity calculations on the basis of 'significance indicators', applied above defined threshold values. These threshold values represent a good surrogate for 'large release', which is defined in a number of EU countries. In addition, the results are placed in the context of in-containment source term limits, for advanced light water reactor designs, as defined by international guidelines. Overall, despite the phenomenological uncertainties, the predicted source terms (both into the containment, and subsequently, into the environment) do not display a high degree of sensitivity to the individual fission product issues addressed in this project. This is due, mainly, to the substantial capacity for the attenuation of airborne fission products by the designed safety provisions and the natural fission product retention mechanisms within the containment

  15. CONTEMPT: computer program for predicting containment pressure-temperature response to a loss-of-coolant accident

    The CONTEMPT code is used by Babcock and Wilcox for containment analysis following a postulated loss of coolant accident. An additional model is described which is used for the calculation of long term post reflood mass and energy releases to the containment that is used for the containment design basis LOCA calculations. These calculations maximize the rate of energy flow to the containment. The mass and energy data are given to the containment designer for use in calculating the containment building design pressure and temperature and in sizing containment heat removal equipment

  16. WSPEEDI (worldwide version of SPEEDI): A computer code system for the prediction of radiological impacts on Japanese due to a nuclear accident in foreign countries

    A computer code system has been developed for near real-time dose assessment during radiological emergencies. The system WSPEEDI, the worldwide version of SPEEDI (System for Prediction of Environmental Emergency Dose Information) aims at predicting the radiological impact on Japanese due to a nuclear accident in foreign countries. WSPEEDI consists of a mass-consistent wind model WSYNOP for large-scale wind fields and a particle random walk model GEARN for atmospheric dispersion and dry and wet deposition of radioactivity. The models are integrated into a computer code system together with a system control software, worldwide geographic database, meteorological data processor and graphic software. The performance of the models has been evaluated using the Chernobyl case with reliable source terms, well-established meteorological data and a comprehensive monitoring database. Furthermore, the response of the system has been examined by near real-time simulations of the European Tracer Experiment (ETEX), carried out over about 2,000 km area in Europe. (author)

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

    Blaise Collin

    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

  18. Modelling the radionuclide contamination of the Black Sea in the result of Chernobyl accident using circulation model and data assimilation

    Assimilation of observations is powerful tool to improve the predictive capabilities of models for radionuclide transport and fate. We describe results of numerical experiments on assimilation of data on radionuclide contamination of the Black Sea in the result of Chernobyl accident in the three-dimensional model of circulation and radionuclide transport THREETOX. Data assimilation in THREETOX can be formulated as procedure that contains two steps: update and forecast. On update step THREETOX is to be run from the time of release or the time of deposition to the time of forecast using updated input from all the data available at this period to be assimilated. On forecast step THREETOX is to be run from forecast time to the end of modeling period using results of update step to produce forecast of radionuclide transport. Two data assimilation methods were used: steady state specific Kalman filter (SSKF) and method iteration of optimal solution (IOS). These methods have been applied to the assimilation observational data of 137Cs concentration in the typhoon surveys for the period June 1986 to September 1990. From results of numerical experiments we conclude, that data assimilation can essentially improve predictive capability of models for radionuclide transport. The IOS was more effective than SSSKF and computational time for this method is small in comparison with general time for calculation. Results from the study will provide a better understanding of the processes of radionuclide transport in the seas. This novel approach is implemented in the EU DSS RODOS for a real-time simulation of the radioactivity transport in the marine environment. (author)

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

    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

  20. Evaluation and Prediction present of radionuclide for surface wipe sample in Emergency Related with Fukushima Nuclear Accident

    Surface wipe samples of aircraft and container from Japan that were exposed to radioactive dust fallout due to Fukushima nuclear accident has been analysed using gamma spectrometry systems. The samples were analysed to determine their contamination levels. The surface of aircraft and container might be exposed to short and long lived fission and activation products. Thus, good evaluations, as well as a reliable and reasonable judgment were needed in order to determine the presence of fission and activation products. A work procedure has been developed to evaluate and predict the presence of fission and activation products in surface wipe samples. Good references, skilled and experienced level in analysis, a well calibrated and validated detector system were the important factors in determining the presence of fission and activation products in surface wipe samples. (author)

  1. Modelling the release of volatile fission product cesium from CANDU fuel under severe accident conditions using artificial neural networks

    An artificial neural network (ANN) model has been developed to predict the release of volatile fission products from CANDU fuel under severe accident conditions. The model was based on data for the release Of 134Cs measured during three annealing experiments (Hot Cell Experiments 1 and 2, or HCE- 1, HCE-2 and Metallurgical Cell Experiment 1, or MCE- 1) at Chalk River Laboratories. These experiments were comprised of a total of 30 separate tests. The ANN established a correlation among 14 separate input variables and predicted the cumulative fractional release for a set of 386 data points drawn from 29 tests to a normalized error, En, of 0.104 and an average absolute error, Eabs, of 0.064. Predictions for a blind validation set (test HCE2-CM6) had an En of 0.064 and an Eabs of 0.054. A methodology is presented for deploying the ANN model by providing the connection weights. Finally, the performance of an ANN model was compared to a fuel oxidation model developed by Lewis et al. and to the U.S. Nuclear Regulatory Commission's CORSOR-M. (author)

  2. A dynamical model of Sayano-Shushenskaya hydropower plant: stability, oscillations, and accident

    Leonov, G. A.; Kuznetsov, N. V.; Solovyeva, E. P.

    2015-01-01

    This work is devoted to the construction and study of a mathematical model of hydropower unit, consisting of synchronous generator, hydraulic turbine, and speed governor. It is motivated by the accident happened on the Sayano-Shushenskaya hydropower plant in 2009 year. Parameters of the Sayano-Shushenskaya hydropower plant were used for modeling the system. Oscillations in zones, which were not recommended for operation, were found. The obtained results are consistent with the full-scale test...

  3. Nonlinear chaotic model for predicting storm surges

    M. Siek

    2010-09-01

    Full Text Available This paper addresses the use of the methods of nonlinear dynamics and chaos theory for building a predictive chaotic model from time series. The chaotic model predictions are made by the adaptive local models based on the dynamical neighbors found in the reconstructed phase space of the observables. We implemented the univariate and multivariate chaotic models with direct and multi-steps prediction techniques and optimized these models using an exhaustive search method. The built models were tested for predicting storm surge dynamics for different stormy conditions in the North Sea, and are compared to neural network models. The results show that the chaotic models can generally provide reliable and accurate short-term storm surge predictions.

  4. Staying Power of Churn Prediction Models

    Risselada, Hans; Verhoef, Peter C.; Bijmolt, Tammo H. A.

    2010-01-01

    In this paper, we study the staying power of various churn prediction models. Staying power is defined as the predictive performance of a model in a number of periods after the estimation period. We examine two methods, logit models and classification trees, both with and without applying a bagging

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

    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

  6. Use of scale models to assess structural response of nuclear shipping containers under accident conditions

    Experimental scale modelling techniques were used to investigate the complex behaviour of truck-type high level waste and spent fuel shipping packaging during severe impact accidents. A series of experiments were conducted with distorted replica scale models fabricated with Type 304 stainless steel and unbonded lead shielding. The models were fabricated to represent typical 1/8, 1/4, and 1/2 linear scaled versions of a full scale prototype unit. Experiments were conducted for 9m (30 ft) free fall accidents onto an essentially unyielding surface with the centre-of-gravity of the model directly over the centre of its bottom plate. The test temperatures were - 40 and 1750C to cover the extreme environmental conditions that this type of packaging may encounter in its normal service life. The inertial loading of the model was controlled during the simulated impact accident by attaching a balsa wood impact limiter to the bottom of the model. Deceleration measurements obtained during the tests were in the range of 1000g. Permanent strain induced in the steel shells was in the range of 0.004 m/m with the largest strain induced at 1750C as expected. Lead slump occurred in all experiments and was in the range of 1 to 3% of the original shielded length. (author)

  7. Modeling control room crews in accident sequence analysis

    Studies of small groups in business, political, and civil aviation environments have found that communication and other group interactions can be critical factors in evaluating group performance. This paper presents a simulation-based model for a nuclear power plant control room crew that treats these interactions as well as operator cognitive behavior. It also provides a rationale for emphasizing the treatment of group interaction, and discusses the current status of the model. (author)

  8. Predictive Modelling and Time: An Experiment in Temporal Archaeological Predictive Models

    David Ebert

    2006-08-01

    Full Text Available One of the most common criticisms of archaeological predictive modelling is that it fails to account for temporal or functional differences in sites. However, a practical solution to temporal or functional predictive modelling has proven to be elusive. This article discusses temporal predictive modelling, focusing on the difficulties of employing temporal variables, then introduces and tests a simple methodology for the implementation of temporal modelling. The temporal models thus created are then compared to a traditional predictive model.

  9. Prediction of thermoplastic failure of a reactor pressure vessel under a postulated core melt accident

    This paper presents the lower head failure calculations performed for a postulated accident scenario in a commercial nuclear power plant. A postulated one inch break in the primary coolant circuit leads to dryout and subsequent meltdown of the core. The reference plant is a pressurized water reactor without penetrations in the reactor vessel lower head. The molten core material accumulates in the lower head, eventually causing failure of the vessel. The analysis investigates flow conditions in the melt pool, temperature evolution in the reactor vessel wall, and structure mechanical evaluation of the vessel under strong thermal loads and a range of internal pressures. The calculations were performed using the ADINA finite element codes. The analysis focusses on the failure processes, time and mode of failure. The most likely mode of failure at low pressure is global rupture due to gradual accumulation of creep strain over a large part of the heated area. In contrast, thermoplasticity becomes important at high pressure or following a pressure spike and can lead to earlier local failure. In situations in which part of the heat load is concentrated over a small area, resulting in a hot spot, local failure occurs, but not until the temperatures are close to the melting point. At low pressure, in particular, the hot spot area remains intact until the structure is molten across more than half of the thickness. (author) 14 figs., 16 refs

  10. Predictive models of radiative neutrino masses

    Julio, J.

    2016-06-01

    We discuss two models of radiative neutrino mass generation. The first model features one-loop Zee model with Z4 symmetry. The second model is the two-loop neutrino mass model with singly- and doubly-charged scalars. These two models fit neutrino oscillation data well and predict some interesting rates for lepton flavor violation processes.

  11. How to Establish Clinical Prediction Models.

    Lee, Yong Ho; Bang, Heejung; Kim, Dae Jung

    2016-03-01

    A clinical prediction model can be applied to several challenging clinical scenarios: screening high-risk individuals for asymptomatic disease, predicting future events such as disease or death, and assisting medical decision-making and health education. Despite the impact of clinical prediction models on practice, prediction modeling is a complex process requiring careful statistical analyses and sound clinical judgement. Although there is no definite consensus on the best methodology for model development and validation, a few recommendations and checklists have been proposed. In this review, we summarize five steps for developing and validating a clinical prediction model: preparation for establishing clinical prediction models; dataset selection; handling variables; model generation; and model evaluation and validation. We also review several studies that detail methods for developing clinical prediction models with comparable examples from real practice. After model development and vigorous validation in relevant settings, possibly with evaluation of utility/usability and fine-tuning, good models can be ready for the use in practice. We anticipate that this framework will revitalize the use of predictive or prognostic research in endocrinology, leading to active applications in real clinical practice. PMID:26996421

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

    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.

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

    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

  14. Probabilistic models for early and continuing radiobiological effects of nuclear power plant accidents

    In this paper the theoretical basis is discussed for specific dose-rate models developed for estimating the risk of death from early and continuing effects of exposure of man to low linear-energy-transfer radiations. Dose-rate models for hematopoietic death and for death from radiation pneumonitis and/or pulmonary fibrosis are provided and are based partly on an assumed Weibull tolerance dose distribution. The dose-rate models were prepared for use by the Nuclear Regulatory Commission in probabilistic analyses of health effects risks associated with potential light-water nuclear reactor accidents. Both exact and approximate solutions are provided. The approximate solutions require less computer time than the exact solutions and therefore may be preferable for complicated radiation accident scenarios

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

    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.

  16. Distributed framework for modeling and reconstruction of nuclear accidents

    Hofman, Radek; Pecha, Petr; Šmídl, Václav

    Vienna : CTBTO, 2013. s. 63-63. [CTBT Science and Technology 2013 Conference. 17.-21.6. 2013, Vienna] R&D Projects: GA MV VG20102013018 Institutional support: RVO:67985556 Keywords : distributed computing * Bayesian filtering * atmospheric dispersion modeling Subject RIV: DG - Athmosphere Sciences, Meteorology http://library.utia.cas.cz/separaty/2013/AS/hoffman-0394249.pdf

  17. Initial VHTR accident scenario classification: models and data.

    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

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

    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

  19. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    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

  20. Accident consequence analysis models applied to licensing process of nuclear installations, radioactive and conventional industries

    The industrial accidents happened in the last years, particularly in the eighty's decade, had contributed in a significant way to call the attention to government authorities, industry and society as a whole, demanding mechanisms for preventing episodes that could affect people's safety and environment quality. Techniques and methods already thoroughly used in the nuclear, aeronautic and war industries were then adapted for performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. Some models for analyzing the consequences of accidents involving fire and explosion, used in the licensing processes of nuclear and radioactive facilities, are presented in this paper. These models have also application in the licensing of conventional industrial facilities. (author)

  1. Modeling of Spray System Operation under Hydrogen and Steam Emissions in NPP Containment during Severe Accident

    Vadim E. Seleznev

    2011-01-01

    Full Text Available The paper describes one of the variants of mathematical models of a fluid dynamics process inside the containment, which occurs in the conditions of operation of spray systems in severe accidents at nuclear power plant. The source of emergency emissions in this case is the leak of the coolant or rupture at full cross-section of the main circulating pipeline in a reactor building. Leak or rupture characteristics define the localization and the temporal law of functioning of a source of emergency emission (or accrued operating of warmed up hydrogen and steam in the containment. Operation of this source at the course of analyzed accident models should be described by the assignment of the relevant Dirichlet boundary conditions. Functioning of the passive autocatalytic recombiners of hydrogen is described in the form of the complex Newton boundary conditions.

  2. Return Predictability, Model Uncertainty, and Robust Investment

    Lukas, Manuel

    2013-01-01

    Stock return predictability is subject to great uncertainty. In this paper we usethe model confidence set approach to quantify uncertainty about expected utilityfrom investment, accounting for potential return predictability. For monthly USdata and six representative return prediction models, we find that confidence setsare very wide, change significantly with the predictor variables, and frequentlyinclude expected utilities for which the investor prefers not to invest. The lattermotivates a ...

  3. Sparse preconditioning for model predictive control

    Knyazev, Andrew; Malyshev, Alexander,

    2015-01-01

    We propose fast O(N) preconditioning, where N is the number of gridpoints on the prediction horizon, for iterative solution of (non)-linear systems appearing in model predictive control methods such as forward-difference Newton-Krylov methods. The Continuation/GMRES method for nonlinear model predictive control, suggested by T. Ohtsuka in 2004, is a specific application of the Newton-Krylov method, which uses the GMRES iterative algorithm to solve a forward difference approximation of the opt...

  4. Meta-analysis of clinical prediction models

    Debray, T.P.A.

    2013-01-01

    The past decades there has been a clear shift from implicit to explicit diagnosis and prognosis. This includes appreciation of clinical -diagnostic and prognostic- prediction models, which is likely to increase with the introduction of fully computerized patient records. Prediction models aim to pro

  5. Unreachable Setpoints in Model Predictive Control

    Rawlings, James B.; Bonné, Dennis; Jørgensen, John Bagterp;

    2008-01-01

    In this work, a new model predictive controller is developed that handles unreachable setpoints better than traditional model predictive control methods. The new controller induces an interesting fast/slow asymmetry in the tracking response of the system. Nominal asymptotic stability of the optimal...

  6. ASTRID model for predicting the radioactive releases to the containment and the environment

    Full text of publication follows: In case of accident affecting a nuclear reactor, it is essential to anticipate the possible development of the situation to efficiently succeed in emergency response actions, i.e. firstly to be early warned, to get sufficient information on the plant and as far as possible, to identify risk extent in terms of potential radioactive release and consequences as a function of time. In this objective, the work of technical teams should be organized and structured in order to assess plant status in a systematic and exhaustive way and to limit the time needed for evaluation. The ASTRID (Assessment of Source Term for Emergency Response based on Installation Data) project consists in developing a methodology of expertise to structure the work of technical teams and to facilitate cross competence communications among EP players and a qualified computer tool that could be commonly used by the European countries to reliably predict source term in case of an accident in a light water reactor, using the information available on the plant. The ASTRID process model is used for the faster than real-time prediction of the radioactivity releases into the containment and further into the environment in case of a severe accident type of an emergency situation in a light water reactor. The graphical user interface offers tools for defining the plant description. The European plants considered as reference plants are the Westinghouse and the French type 4-loop PWR, the French type 3-loop PWR, the Westinghouse type 2-loop PWR, the German type 4-loop Konvoi PWR, the European 4-loop EPR, the VVER-440 type PWR, the VVER-1000 type PWR, the ASEA type internal pump BWR and the ASEA type external loop and pump BWR. During the development phase the inputs have been created into the database of the analysis tools. By the user interface different optional tasks are coordinated including accident diagnosis, source term pre-calculation, accident simulation, and

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

    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

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

    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

  9. Verification of fuel-coolant interaction model for severe accident simulations

    Results of recent verification studies of VAPEX-M module intended for the calculation of fuel-coolant interaction (FCI) are presented. The mathematical model and correlations for the main physical processes are described. Comparisons of calculated results with three series of FCI experiments (MAGICO-2000, QUEOS, FARO) are presented. It is shown that the main features of melt-water interaction are reproduced by VAPEX-M with reasonable accuracy, which makes the module a useful tool for severe accident analysis. (author)

  10. Intracardiac therapy following emergency thoracotomy in the accident and emergency department: an experimental model.

    Moulton, C; Pennycook, A; Crawford, R

    1992-01-01

    For a select group of patients with penetrating chest trauma, immediate thoracotomy in the accident and emergency department offers the only chance of survival. Foley catheters have been used to achieve haemostasis in cardiac wounds but are not widely used for intracardiac fluid and drug administration during resuscitation. In an anatomical model designed to assess this procedure an average flow rate of 275 ml min-1 was achieved. The equipment required is readily available and easily assembled.

  11. Modeling uncertainty: Predictive accuracy as a proxy for predictive confidence

    Rich, Robert; Tracy, Joseph

    2003-01-01

    This paper evaluates current strategies for the empirical modeling of forecast behavior. In particular, we focus on the reliability of using proxies from time series models of heteroskedasticity to describe changes in predictive confidence. We address this issue by examining the relationship between ex post forecast errors and ex ante measures of forecast uncertainty from data on inflation forecasts from the Survey of Professional Forecasters. The results provide little evidence of a strong l...

  12. Modelling of whole-core release of fission products in PWR core melt accidents: Chapter 13

    The computer code FISREL combines the thermal history of a reactor core with experimentally-based release rate constants to calculate whole-core release histories of fission products in PWR core melt accidents. Predictions of the code for releases of volatile fission products during large-break, small-break and transient initiated sequences are presented, and the sensitivities of results to input data examined. A preliminary assessment of the limitations imposed by mass transport on release of vaporized materials in high pressure sequences is given, and the implications of the results for primary system transport are discussed

  13. Risk assessment and remedial policy evaluation using predictive modeling

    As a result of nuclear industry operation and accidents, large areas of natural ecosystems have been contaminated by radionuclides and toxic metals. Extensive societal pressure has been exerted to decrease the radiation dose to the population and to the environment. Thus, in making abatement and remediation policy decisions, not only economic costs but also human and environmental risk assessments are desired. This paper introduces a general framework for risk assessment and remedial policy evaluation using predictive modeling. Ecological risk assessment requires evaluation of the radionuclide distribution in ecosystems. The FORESTPATH model is used for predicting the radionuclide fate in forest compartments after deposition as well as for evaluating the efficiency of remedial policies. Time of intervention and radionuclide deposition profile was predicted as being crucial for the remediation efficiency. Risk assessment conducted for a critical group of forest users in Belarus shows that consumption of forest products (berries and mushrooms) leads to about 0.004% risk of a fatal cancer annually. Cost-benefit analysis for forest cleanup suggests that complete removal of organic layer is too expensive for application in Belarus and a better methodology is required. In conclusion, FORESTPATH modeling framework could have wide applications in environmental remediation of radionuclides and toxic metals as well as in dose reconstruction and, risk-assessment

  14. A Theoretical Model for the Prediction of Siphon Breaking Phenomenon

    A siphon phenomenon or siphoning often refers to the movement of liquid from a higher elevation to a lower one through a tube in an inverted U shape (whose top is typically located above the liquid surface) under the action of gravity, and has been used in a variety of reallife applications such as a toilet bowl and a Greedy cup. However, liquid drainage due to siphoning sometimes needs to be prevented. For example, a siphon breaker, which is designed to limit the siphon effect by allowing the gas entrainment into a siphon line, is installed in order to maintain the pool water level above the reactor core when a loss of coolant accident (LOCA) occurs in an open-pool type research reactor. In this paper, we develop a theoretical model to predict the siphon breaking phenomenon. In this paper, a theoretical model to predict the siphon breaking phenomenon is developed. It is shown that the present model predicts well the fundamental features of the siphon breaking phenomenon and undershooting height

  15. A Theoretical Model for the Prediction of Siphon Breaking Phenomenon

    Bae, Youngmin; Kim, Young-In; Seo, Jae-Kwang; Kim, Keung Koo; Yoon, Juhyeon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2014-10-15

    A siphon phenomenon or siphoning often refers to the movement of liquid from a higher elevation to a lower one through a tube in an inverted U shape (whose top is typically located above the liquid surface) under the action of gravity, and has been used in a variety of reallife applications such as a toilet bowl and a Greedy cup. However, liquid drainage due to siphoning sometimes needs to be prevented. For example, a siphon breaker, which is designed to limit the siphon effect by allowing the gas entrainment into a siphon line, is installed in order to maintain the pool water level above the reactor core when a loss of coolant accident (LOCA) occurs in an open-pool type research reactor. In this paper, we develop a theoretical model to predict the siphon breaking phenomenon. In this paper, a theoretical model to predict the siphon breaking phenomenon is developed. It is shown that the present model predicts well the fundamental features of the siphon breaking phenomenon and undershooting height.

  16. Development and verification of models for cladding oxidation in the early stages of a severe accident

    Models have been developed for fuel cladding oxidation in the early stages of a severe accident. The models take account of the crucibilization effect of the zirconium dioxide layer on preventing the Zircaloy melt from flowing down, inner surface oxidation, and the steam starvation effect. The models have been included in the core thermal-hydraulic code SEFDAN. The SEFDAN code has been applied to analyses of severe fuel damage tests in the PBF and NRU facilities, and the TMI-2 accident. The calculated results are in good agreement with the measured or observed results. The analyses indicate that the fuel cladding temperature would have reached the melting point of zirconium dioxide in the PBF.SFD tests and in the TMI-2 accident, and that the fuel temperature would have reached the melting point of uranium dioxide in the PBF.SFD scoping test and TMI-2. In leading fuel rods to such high temperatures, the crucibilization effect of the zirconium dioxide layer plays an essential role, because it retains the molten Zircaloy and sustains the zirconium-water reaction of the molten Zircaloy. In addition, the significant role of the inner surface oxidation on temperature escalation and hydrogen generation rate has been revealed in the analyses of the NRU.FLHT-2 test. (author). 16 refs, 7 figs

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

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

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

    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

  19. Combinatorial Modelling and Learning with Prediction Markets

    Hu, Jinli

    2012-01-01

    Combining models in appropriate ways to achieve high performance is commonly seen in machine learning fields today. Although a large amount of combinatorial models have been created, little attention is drawn to the commons in different models and their connections. A general modelling technique is thus worth studying to understand model combination deeply and shed light on creating new models. Prediction markets show a promise of becoming such a generic, flexible combinatorial model. By reviewing on several popular combinatorial models and prediction market models, this paper aims to show how the market models can generalise different combinatorial stuctures and how they implement these popular combinatorial models in specific conditions. Besides, we will see among different market models, Storkey's \\emph{Machine Learning Markets} provide more fundamental, generic modelling mechanisms than the others, and it has a significant appeal for both theoretical study and application.

  20. Modeling of hot tensile and short-term creep strength for LWR piping materials under severe accident conditions

    The analytical study on severe accident shows the possibility of the reactor coolant system (RCS) piping failure before reactor pressure vessel failure under the high primary pressure sequence at pressurized water reactors. The establishment of the high-temperature strength model of the realistic RCS piping materials is important in order to predict precisely the accident progression and to evaluate the piping behavior with small uncertainties. Based on material testing, the 0.2% proof stress and the ultimate tensile strength above 800degC were given by the equations of second degree as a function of the reciprocal absolute temperature considering the strength increase due to fine precipitates for the piping materials. The piping materials include type 316 stainless steel, type 316 stainless steel of nuclear grade, CF8M cast duplex stainless steel and STS410 carbon steel. Also the short-term creep rupture time and the minimum creep rate at high-temperature were given by the modified Norton's Law as a function of stress and temperature considering the effect of the precipitation formation and resolution on the creep strength. The present modified Norton's Law gives better results than the conventional Larson-Miller method. Correlating the creep data (the applied stress versus the minimum creep rate) with the tensile data (the 0.2% proof stress or the ultimate tensile strength versus the strain rate), it was found that the dynamic recrystallization significantly occurred at high-temperature. (author)

  1. Real Time Predictions of Transport, Dispersion and Deposition from Nuclear Accidents

    1999-01-01

    Proceedings of the NATO Advanced Research Workshop on Large Scale Computations in Air Pollution Modelling, Sofia, Bulgaria, 6-10 July 1998......Proceedings of the NATO Advanced Research Workshop on Large Scale Computations in Air Pollution Modelling, Sofia, Bulgaria, 6-10 July 1998...

  2. Model predictive control classical, robust and stochastic

    Kouvaritakis, Basil

    2016-01-01

    For the first time, a textbook that brings together classical predictive control with treatment of up-to-date robust and stochastic techniques. Model Predictive Control describes the development of tractable algorithms for uncertain, stochastic, constrained systems. The starting point is classical predictive control and the appropriate formulation of performance objectives and constraints to provide guarantees of closed-loop stability and performance. Moving on to robust predictive control, the text explains how similar guarantees may be obtained for cases in which the model describing the system dynamics is subject to additive disturbances and parametric uncertainties. Open- and closed-loop optimization are considered and the state of the art in computationally tractable methods based on uncertainty tubes presented for systems with additive model uncertainty. Finally, the tube framework is also applied to model predictive control problems involving hard or probabilistic constraints for the cases of multiplic...

  3. Modelling and simulation of the radioactive release during the Fukushima accident

    The nuclear accident of Fukushima raised again the discussion how operating companies and authorities can react in such a case of emergency. This paper investigates from a scientific perspective, how the tools and measures introduced in the last years, in particular the simulation models, can be applied to assist the decision makers. To this end, the simulation system ABR, which calculates the dispersion of radioactive particles and the resulting radiological exposure in general, is introduced. Especially, it is shown how the ABR system has been adopted to simulate the Fukushima accident. The assumptions that were made to determine the most important data like the source term and weather conditions are described and the simulation results obtained by the ABR system are discussed. (orig.)

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

    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

  5. Disease Models for Event Prediction

    Corley, Courtney D.; Pullum, Laura

    2013-01-01

    Objective The objective of this manuscript is to present a systematic review of biosurveillance models that operate on select agents and can forecast the occurrence of a disease event. Introduction One of the primary goals of this research was to characterize the viability of biosurveillance models to provide operationally relevant information to decision makers, in order to identify areas for future research. Two critical characteristics differentiate this work from other infectious disease ...

  6. Modeling in fast dynamics of accidents in the primary circuit of PWR type reactors

    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)

  7. Validation of advanced NSSS simulator model for loss-of-coolant accidents

    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.

  8. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume II

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperature and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters are required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtle number for saturated liquid, saturated vapour, subcooled liquid of superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its structure

  9. Microcomputer based program for predicting heat transfer under reactor accident conditions. Volume I

    A microcomputer based program called Heat Transfer Prediction Software (HTPS) has been developed. It calculates the heat transfer for the tube and bundle geometries for steady state and transient conditions. This program is capable of providing the best estimated of the hot pin temperatures during slow transients for 37- and 28-element CANDU type fuel bundles. The program is designed for an IBM-PC AT/XT (or IBM-PC compatible computer) equipped with a Math Co-processor. The following input parameters are required: pressure, mass flux, hydraulic diameter, and quality. For the steady state case, the critical heat flux (CHF), the critical heat flux temperature, the minimum film boiling temperature, and the minimum film boiling heat flux are the primary outputs. With either the surface heat flux or wall temperature specified, the program determines the heat transfer regime and calculates the surface heat flux, wall temperatures and heat transfer coefficient. For the slow transient case, the pressure, mass flux, quality, and volumetric heat generation rate are the time dependent input parameters required to calculate the hot pin sheath temperatures and surface heat fluxes. A simple routine for generating properties has been developed for light water to support the above program. It contains correlations that have been verified for pressures ranging from 0.6kPa to 30 MPa, and temperatures up to 1100 degrees Celcius. The thermodynamic and transport properties that can be generated from this routine are: density, specific volume, enthalpy, specific heat capacity, conductivity, viscosity, surface tension and Prandtl number for saturated liquid, saturated vapour, subcooled liquid for superheated vapour. A software for predicting flow regime has also been developed. It determines the flow pattern at specific flow conditions, and provides a correction factor for calculating the CHF during partially stratified horizontal flow. The technical bases for the program and its

  10. Accidents - Chernobyl accident

    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.)

  11. Energy based prediction models for building acoustics

    Brunskog, Jonas

    2012-01-01

    In order to reach robust and simplified yet accurate prediction models, energy based principle are commonly used in many fields of acoustics, especially in building acoustics. This includes simple energy flow models, the framework of statistical energy analysis (SEA) as well as more elaborated...... principles as, e.g., wave intensity analysis (WIA). The European standards for building acoustic predictions, the EN 12354 series, are based on energy flow and SEA principles. In the present paper, different energy based prediction models are discussed and critically reviewed. Special attention is placed on...

  12. Massive Predictive Modeling using Oracle R Enterprise

    CERN. Geneva

    2014-01-01

    R is fast becoming the lingua franca for analyzing data via statistics, visualization, and predictive analytics. For enterprise-scale data, R users have three main concerns: scalability, performance, and production deployment. Oracle's R-based technologies - Oracle R Distribution, Oracle R Enterprise, Oracle R Connector for Hadoop, and the R package ROracle - address these concerns. In this talk, we introduce Oracle's R technologies, highlighting how each enables R users to achieve scalability and performance while making production deployment of R results a natural outcome of the data analyst/scientist efforts. The focus then turns to Oracle R Enterprise with code examples using the transparency layer and embedded R execution, targeting massive predictive modeling. One goal behind massive predictive modeling is to build models per entity, such as customers, zip codes, simulations, in an effort to understand behavior and tailor predictions at the entity level. Predictions...

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

    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

  14. Lung Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing lung cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  15. Prostate Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing prostate cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  16. Breast Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing breast cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  17. Optimal Prediction in Loglinear Models

    K.J. van Garderen

    2001-01-01

    This paper introduces a Laplace inversion technique for deriving unbiased predictors in exponential families. This general technique is applied to derive the exact optimal unbiased predictor in loglinear models with Gaussian disturbances under quadratic loss. An exact unbiased estimator for its vari

  18. Predictive Modeling in Adult Education

    Lindner, Charles L.

    2011-01-01

    The current economic crisis, a growing workforce, the increasing lifespan of workers, and demanding, complex jobs have made organizations highly selective in employee recruitment and retention. It is therefore important, to the adult educator, to develop models of learning that better prepare adult learners for the workplace. The purpose of…

  19. Ovarian Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing ovarian cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  20. Cervical Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing cervical cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  1. Liver Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing liver cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  2. Pancreatic Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing pancreatic cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  3. Colorectal Cancer Risk Prediction Models

    Developing statistical models that estimate the probability of developing colorectal cancer over a defined period of time will help clinicians identify individuals at higher risk of specific cancers, allowing for earlier or more frequent screening and counseling of behavioral changes to decrease risk.

  4. Traffic Congestion and Accidents

    Schrage, Andrea

    2006-01-01

    Obstructions caused by accidents can trigger or exacerbate traffic congestion. This paper derives the efficient traffic pattern for a rush hour with congestion and accidents and the corresponding road toll. Compared to the model without accidents, where the toll equals external costs imposed on drivers using the road at the same time, a new insight arises: An optimal toll also internalizes the expected increase in future congestion costs. Since accidents affect more drivers if traffic volumes...

  5. Modeling and Prediction Using Stochastic Differential Equations

    Juhl, Rune; Møller, Jan Kloppenborg; Jørgensen, John Bagterp;

    2016-01-01

    Pharmacokinetic/pharmakodynamic (PK/PD) modeling for a single subject is most often performed using nonlinear models based on deterministic ordinary differential equations (ODEs), and the variation between subjects in a population of subjects is described using a population (mixed effects) setup...... deterministic and can predict the future perfectly. A more realistic approach would be to allow for randomness in the model due to e.g., the model be too simple or errors in input. We describe a modeling and prediction setup which better reflects reality and suggests stochastic differential equations (SDEs) for...

  6. DC Motor Control Predictive Models

    Ravinesh Singh; Godfrey C. Onwubolu; Krishnileshwar Singh; Ritnesh Ram

    2006-01-01

    DC motor speed and position controls are fundamental in vehicles in general and robotics in particular. This study presents a mathematical model for correlating the interactions of some DC motor control parameters such as duty cycle, terminal voltage, frequency and load on some responses such as output current, voltage and speed by means of response surface methodology. For this exercise, a five-level full factorial design was chosen for experimentation using a peripheral interface controller...

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

    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

  8. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; ''Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents''

  9. Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents

    Siefken, Larry James

    1999-02-01

    Preliminary designs are described for models of hydrogen and oxygen uptake in fuel rod cladding during severe accidents. Calculation of the uptake involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the clad-ding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental results are presented that show a rapid uptake of hydrogen in the event of dissolution of the oxide layer and a rapid release of hydrogen in the event of cracking of the oxide layer. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for hydrogen and oxygen uptake and cladding embrittlement into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the proposed models on the calculated behavior of fuel rods in severe accident conditions. This report is a revision and reissue of the report entitled; "Preliminary Design Report for Modeling of Hydrogen Uptake in Fuel Rod Cladding During Severe Accidents."

  10. A Course in... Model Predictive Control.

    Arkun, Yaman; And Others

    1988-01-01

    Describes a graduate engineering course which specializes in model predictive control. Lists course outline and scope. Discusses some specific topics and teaching methods. Suggests final projects for the students. (MVL)

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

    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.)

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

    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.

  13. Equivalency and unbiasedness of grey prediction models

    Bo Zeng; Chuan Li; Guo Chen; Xianjun Long

    2015-01-01

    In order to deeply research the structure discrepancy and modeling mechanism among different grey prediction mo-dels, the equivalence and unbiasedness of grey prediction mo-dels are analyzed and verified. The results show that al the grey prediction models that are strictly derived from x(0)(k) +az(1)(k) = b have the identical model structure and simulation precision. Moreover, the unbiased simulation for the homoge-neous exponential sequence can be accomplished. However, the models derived from dx(1)/dt+ax(1) =b are only close to those derived from x(0)(k)+az(1)(k)=b provided that|a|has to satisfy|a| < 0.1; neither could the unbiased simulation for the homoge-neous exponential sequence be achieved. The above conclusions are proved and verified through some theorems and examples.

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

    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.

  15. Radiocesium contamination in a submediterranean semi-natural ecosystem following the Chernobyl accident: Measurements and models

    Radiocesium dynamics in a Quercus conferta Kit ecosystem in Northern Greece have been extensively studied over the years 1993-1995. Radiocesium distribution in the different parts of the ecosystem was measured. A total 137Cs inventory of 243±66 MBq ha-1 due to the Chernobyl accident was measured in all parts of the ecosystem. Almost 90% of this inventory is still in the upper layers of the soil and the forest floor. In particular 13.4% is in the forest floor, 52.6% in the Ah horizon, and 23.4% in the upper 5 cm of the soil. Only 2.2% of this inventory is in the above ground biomass. The mean total 137Cs deposited on the forest floor from the above ground biomass is 0.18 MBq ha-1 y-1. Cesium leaching from the forest floor is negligible. The radiocesium distribution in soil is fixed and in equilibrium, at least since 1993. Most of radiocesium is not available for migration. Cesium migration in soil was modeled by (a) an open-quotes equivalent diffusionclose quotes model with different initial conditions and (b) a open-quotes compartmentclose quotes model derived from a diffusion-advection model. A compartment model for the contamination of living biomass is proposed. The total absorbed dose rate in air as well as the contribution due to 137Cs from the Chernobyl accident was determined inside the forest, by in-situ gamma spectrometry. 32 refs., 10 figs., 7 tabs

  16. Return Predictability, Model Uncertainty, and Robust Investment

    Lukas, Manuel

    Stock return predictability is subject to great uncertainty. In this paper we use the model confidence set approach to quantify uncertainty about expected utility from investment, accounting for potential return predictability. For monthly US data and six representative return prediction models, we...... find that confidence sets are very wide, change significantly with the predictor variables, and frequently include expected utilities for which the investor prefers not to invest. The latter motivates a robust investment strategy maximizing the minimal element of the confidence set. The robust investor...

  17. Predictive technology model for robust nanoelectronic design

    Cao, Yu

    2011-01-01

    Predictive Technology Model for Robust Nanoelectronic Design explains many of the technical mysteries behind the Predictive Technology Model (PTM) that has been adopted worldwide in explorative design research. Through physical derivation and technology extrapolation, PTM is the de-factor device model used in electronic design. This work explains the systematic model development and provides a guide to robust design practice in the presence of variability and reliability issues. Having interacted with multiple leading semiconductor companies and university research teams, the author brings a s

  18. Accuracy assessment of landslide prediction models

    The increasing population and expansion of settlements over hilly areas has greatly increased the impact of natural disasters such as landslide. Therefore, it is important to developed models which could accurately predict landslide hazard zones. Over the years, various techniques and models have been developed to predict landslide hazard zones. The aim of this paper is to access the accuracy of landslide prediction models developed by the authors. The methodology involved the selection of study area, data acquisition, data processing and model development and also data analysis. The development of these models are based on nine different landslide inducing parameters i.e. slope, land use, lithology, soil properties, geomorphology, flow accumulation, aspect, proximity to river and proximity to road. Rank sum, rating, pairwise comparison and AHP techniques are used to determine the weights for each of the parameters used. Four (4) different models which consider different parameter combinations are developed by the authors. Results obtained are compared to landslide history and accuracies for Model 1, Model 2, Model 3 and Model 4 are 66.7, 66.7%, 60% and 22.9% respectively. From the results, rank sum, rating and pairwise comparison can be useful techniques to predict landslide hazard zones

  19. Estimation of the time-dependent radioactive source-term from the Fukushima nuclear power plant accident using atmospheric transport modelling

    Schoeppner, M.; Plastino, W.; Budano, A.; De Vincenzi, M.; Ruggieri, F.

    2012-04-01

    Several nuclear reactors at the Fukushima Dai-ichi power plant have been severely damaged from the Tōhoku earthquake and the subsequent tsunami in March 2011. Due to the extremely difficult on-site situation it has been not been possible to directly determine the emissions of radioactive material. However, during the following days and weeks radionuclides of 137-Caesium and 131-Iodine (amongst others) were detected at monitoring stations throughout the world. Atmospheric transport models are able to simulate the worldwide dispersion of particles accordant to location, time and meteorological conditions following the release. The Lagrangian atmospheric transport model Flexpart is used by many authorities and has been proven to make valid predictions in this regard. The Flexpart software has first has been ported to a local cluster computer at the Grid Lab of INFN and Department of Physics of University of Roma Tre (Rome, Italy) and subsequently also to the European Mediterranean Grid (EUMEDGRID). Due to this computing power being available it has been possible to simulate the transport of particles originating from the Fukushima Dai-ichi plant site. Using the time series of the sampled concentration data and the assumption that the Fukushima accident was the only source of these radionuclides, it has been possible to estimate the time-dependent source-term for fourteen days following the accident using the atmospheric transport model. A reasonable agreement has been obtained between the modelling results and the estimated radionuclide release rates from the Fukushima accident.

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

    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

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

    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)

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

    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)

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

    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.

  4. Modelling of Nuclear Fuel Under Accident Conditions by Means of Transuranus

    The TRANSURANUS fuel performance code, which is developed at the JRC-ITU and in collaboration with many partner institutes since more than three decades, has been adapted in order to be able to simulate design basis accident (DBA) conditions. In a first step, the developments and associated validation work will be summarised for LOCA conditions. This part includes modifications in the model for large strains, for the crystallographic phase transition in Zircaloy, and for burst release and large cladding deformations. In a second step, the ongoing work for simulations of RIA conditions will be outlined that include the model for the plenum temperature, along with the separate effect studies and detailed model developments made in parallel by means of multi-scale and multi-physics tools for the high burnup structure. Finally, the perspectives of model developments and needs for further verification and validation in the frame of international benchmark exercises dedicated to DBA simulations and the first phase of a severe accident, i.e. when the cylindrical fuel rod geometry is preserved, will be presented for discussion. (author)

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

    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

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

    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. Prediction of PARP Inhibition with Proteochemometric Modelling and Conformal Prediction.

    Cortés-Ciriano, Isidro; Bender, Andreas; Malliavin, Thérèse

    2015-06-01

    Poly(ADP-ribose) polymerases (PARPs) play a key role in DNA damage repair. PARP inhibitors act as chemo- and radio- sensitizers and thus potentiate the cytotoxicity of DNA damaging agents. Although PARP inhibitors are currently investigated as chemotherapeutic agents, their cross-reactivity with other members of the PARP family remains unclear. Here, we apply Proteochemometric Modelling (PCM) to model the activity of 181 compounds on 12 human PARPs. We demonstrate that PCM (R0 (2) test =0.65-0.69; RMSEtest =0.95-1.01 °C) displays higher performance on the test set (interpolation) than Family QSAR and Family QSAM (Tukey's HSD, α 0.05), and outperforms Inductive Transfer knowledge among targets (Tukey's HSD, α 0.05). We benchmark the predictive signal of 8 amino acid and 11 full-protein sequence descriptors, obtaining that all of them (except for SOCN) perform at the same level of statistical significance (Tukey's HSD, α 0.05). The extrapolation power of PCM to new compounds (RMSE=1.02±0.80 °C) and targets (RMSE=1.03±0.50 °C) is comparable to interpolation, although the extrapolation ability is not uniform across the chemical and the target space. For this reason, we also provide confidence intervals calculated with conformal prediction. In addition, we present the R package conformal, which permits the calculation of confidence intervals for regression and classification caret models. PMID:27490382

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

    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

  9. Application of transient ignition model to multi-canister (MCO) accident analysis

    The potential for ignition of spent nuclear fuel in a Multi-Canister Overpack (MCO) is examined. A transient model is applied to calculate the highest ambient gas temperature outside an MCO wall tube or shipping cask for which a stable temperature condition exists. This integral analysis couples reaction kinetics with a description of the MCO configuration, heat and mass transfer, and fission product phenomena. It thereby allows ignition theory to be applied to various complex scenarios, including MCO water loss accidents and dry MCO air ingression

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

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

    1988-07-01

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

  11. Modelling Validation of Transients and Initial Phase of Accident Scenarios for Sodium Fast Reactors

    Physical phenomena are presented being of importance in case of transients and / or initial phases of severe accidents in Sodium-cooled Fast Reactors. The CABRI-programmes provided experimental data being characteristic for the physical phenomena and providing information to validate models and parameters used in theoretical simulations. Results of post irradiation examination (PIE), post test examination (PTE) and measurements performed during the experimental tests are presented for transient overpower (TOP), transient undercooling overpower (TUCOP), loss-of-flow tests (LOF) and slow power ramps. (author)

  12. Updating and testing of a PWR model for the Modular Accident Analysis Programe MAAP5

    Marcos Delgado, Elisabet

    2013-01-01

    The present Master’s Thesis is part of the Master’s degree in Nuclear Engineering of the Universitat Politècnica de Catalunya and the ENDESA Escuela de Energía, and it was developed during the internship in a Spanish Pressurized Water Reactor (PWR). The objective of the project is to update and test the nuclear plant model used for the Safety Analysis department which belongs to the Licensing Department mainly for Severe Accidents phenomenology studies to prepare for and respond to emergen...

  13. Mathematical model for predicting human vertebral fracture

    Benedict, J. V.

    1973-01-01

    Mathematical model has been constructed to predict dynamic response of tapered, curved beam columns in as much as human spine closely resembles this form. Model takes into consideration effects of impact force, mass distribution, and material properties. Solutions were verified by dynamic tests on curved, tapered, elastic polyethylene beam.

  14. Predictions of nuclear masses in different models

    The modern version of the liquid-drop model is compared to the macroscopic Thomas-Fermi (TF) energy and the macroscopic part of the binding energy evaluated within the Hartree-Fock-Bogoliubov theory with the Gogny force and the relativistic mean field theory. The limits of nuclear stability predicted by these models are discussed. (author)

  15. Modelling Chemical Reasoning to Predict Reactions

    Segler, Marwin H S

    2016-01-01

    The ability to reason beyond established knowledge allows Organic Chemists to solve synthetic problems and to invent novel transformations. Here, we propose a model which mimics chemical reasoning and formalises reaction prediction as finding missing links in a knowledge graph. We have constructed a knowledge graph containing 14.4 million molecules and 8.2 million binary reactions, which represents the bulk of all chemical reactions ever published in the scientific literature. Our model outperforms a rule-based expert system in the reaction prediction task for 180,000 randomly selected binary reactions. We show that our data-driven model generalises even beyond known reaction types, and is thus capable of effectively (re-) discovering novel transformations (even including transition-metal catalysed reactions). Our model enables computers to infer hypotheses about reactivity and reactions by only considering the intrinsic local structure of the graph, and because each single reaction prediction is typically ac...

  16. Preliminary assessment of the impact of candidate accident-tolerant fuels/cladding on the predicted reactor behaviour at normal operating conditions and under DB (LOCA and RIA) and BDB (STSBO and LTSBO) accident conditions

    Currently, the United States Department of Energy (DOE) has initiated the study of advanced accident-tolerant fuel/cladding (ATF) configurations that exhibit 1) slower reaction kinetics with steam, 2) lower enthalpy of oxidation, 3) less susceptibility to unfavourable core material interactions, and 4) provision of additional barriers to fission product release. Whenever changes, whether minor or major, are made to commercial NPP fuel/cladding systems; then the effect of these changes must be evaluated on all phases of the fuel/cladding lifetime (from fabrication through operation through eventual storage and reprocessing). This presentation focuses on preliminary assessments of several potential ATFs on the impact of these materials on predicted reactor behaviour 1) at normal operating conditions, 2) under postulated design basis (DB) accidents (LOCAs and RIAs), and 3) under beyond design basis (BDB) accident conditions [for short- and long-term station blackouts(SBO)]. These preliminary reactor response predictions are compared against the responses of UO2/Zr cores. For the ATFs evaluated, during normal operation, the most significant features are much lower fuel centerline temperatures and fission gas releases; and for LOCAs the peak cladding temperatures are lower with significantly lower hydrogen generation rates and for a RIA the ATF ejected worth is very similar to the UO2 ejected worth. The use of higher melting/lower hydrogen producing core components (ATFs) will not preclude a BDB accident. Without core cooling the severe accident will march-on; however, the ATFs do allow an increase in margin (time) to initiation of core component degradation - although this may be measured in minutes rather than hours. The ATF core responses (with oxidation kinetics about two orders of magnitude lower than that for Zr) are nearly the same as for components with no oxidation (for a STSBO, the increased time to vessel dry-out is approximately 4.5 hours). There is a need

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

    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.

  18. Development of simplified evaluation models for the first power peak during a criticality accident and its verification by the TRACE code simulated results based on CRAC experimental data

    In a reprocessing facility or a part of uranium fuel manufacturing facility where nuclear fuel solution is processed, one could frequently observe a series of power peaks with the first highest right after a criticality accident. The criticality alarm system (CAS) is designed to detect the first power peak and immediately warn workers around the reacting material by any means such as sounding alarms. Consequently, exposure of the workers could be minimized by an immediate and effective evacuation. Therefore in the design and installation of CAS, it is necessary to estimate the magnitude of the first power peak and to set up the threshold point for CAS initiating alarm. Furthermore, it is necessary to estimate the potential level of accidental exposure of workers so as to decide whether or not it is appropriate to install CAS for any compartment. In this report, simplified evaluation models to estimate the minimum scale of the first power peak and the released energy during a criticality accident are derived only by theoretical consideration for use in the design of CAS to set up the threshold point triggering the alarm signal. Other simplified evaluation models are in the same way derived to estimate the maximum scale of the first power peak and the released energy and to predict possible exposure level of workers to be used to judge the appropriateness of CAS installation. These evaluation models are shown to have adequate margin in predicting the minimum and maximum scale of criticality accidents by comparing their results with French CRAC experiment data. Furthermore, comparison of the maximum scale of the first power peak simplified evaluation, has been made with simulated results by the TRACE code based on the extrapolated conditions predicted by the CRAC experiment data to verify the effectiveness of the derived evaluation models

  19. Evaluation of CASP8 model quality predictions

    Cozzetto, Domenico

    2009-01-01

    The model quality assessment problem consists in the a priori estimation of the overall and per-residue accuracy of protein structure predictions. Over the past years, a number of methods have been developed to address this issue and CASP established a prediction category to evaluate their performance in 2006. In 2008 the experiment was repeated and its results are reported here. Participants were invited to infer the correctness of the protein models submitted by the registered automatic servers. Estimates could apply to both whole models and individual amino acids. Groups involved in the tertiary structure prediction categories were also asked to assign local error estimates to each predicted residue in their own models and their results are also discussed here. The correlation between the predicted and observed correctness measures was the basis of the assessment of the results. We observe that consensus-based methods still perform significantly better than those accepting single models, similarly to what was concluded in the previous edition of the experiment. © 2009 WILEY-LISS, INC.

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

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

  1. RaCon: a software tool serving to predict radiological consequences of various types of accident in support of emergency management and radiation monitoring management

    The RaCon software system, developed by the Nuclear Research Institute Rez, is described and its application when addressing various tasks in the domain of radiation accidents and nuclear safety (accidents at nuclear facilities, transport of radioactive material, terrorist attacks) are outlined. RaCon is intended for the prediction and evaluation of radiological consequences to population and rescue teams and for optimization of monitoring actions. The system provides support to emergency management when evaluating and devising actions to mitigate the consequences of radiation accidents. The deployment of RaCon within the system of radiation monitoring by mobile emergency teams or remote controlled UAV is an important application. Based on a prediction of the radiological situation, RaCon facilitates decision-making and control of the radiation monitoring system, and in turn, refines the prediction based on observed values. Furthermore, the system can perform simulations of evacuation patterns at the Dukovany NPP and at schools in the vicinity of the power plant and can provide support to emergency management should any such situation arise. (orig.)

  2. Efficient particle continuation model predictive control

    Knyazev, Andrew; Malyshev, Alexander,

    2015-01-01

    Continuation model predictive control (MPC), introduced by T. Ohtsuka in 2004, uses Krylov-Newton approaches to solve MPC optimization and is suitable for nonlinear and minimum time problems. We suggest particle continuation MPC in the case, where the system dynamics or constraints can discretely change on-line. We propose an algorithm for on-line controller implementation of continuation MPC for ensembles of predictions corresponding to various anticipated changes and demonstrate its numeric...

  3. Genetic models of homosexuality: generating testable predictions

    Gavrilets, Sergey; Rice, William R.

    2006-01-01

    Homosexuality is a common occurrence in humans and other species, yet its genetic and evolutionary basis is poorly understood. Here, we formulate and study a series of simple mathematical models for the purpose of predicting empirical patterns that can be used to determine the form of selection that leads to polymorphism of genes influencing homosexuality. Specifically, we develop theory to make contrasting predictions about the genetic characteristics of genes influencing homosexuality inclu...

  4. Models for Predictive Railway Traffic Management

    Kecman, P.

    2014-01-01

    The potential growth in transport demand in the next decade and beyond requires a change from reactive to proactive traffic control to maintain and improve the reliability of railway traffic. In order to enable an anticipative approach to traffic management, it is necessary to develop the tools for monitoring, prediction and optimisation of the traffic operations. This thesis presents the models that can be used as components for a decision support system for predictive traffic management.

  5. 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

    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.

  6. Containment Evaluation under Severe Accidents (CESA): synthesis of the predictive calculations and analysis of the first experimental results obtained on the Civaux mock-up

    In 1996, EDF decided to build a containment model at the scale 1:3, the MAEVA mock-up, in order to check and study the behaviour of a pre-stressed concrete containment vessel without a liner in terms of mechanical strength and leaktightness, for loadings corresponding to its design and beyond design conditions. In parallel with the construction and testing of the mock-up, a cost-shared R and D action supported by the European Union, the CESA project, is dealing with quantification of leak rates through concrete cracks and porosity, predictive calculations of the behaviour of the mock-up and analysis of the experimental results. In this paper, we propose a synthesis of the main theoretical and experimental results, obtained after 2.5 years. It should however be noted that, due to some unexpected delays in the experimental programme, quite natural with such a huge and unique experimental set-up, only the design-basis accident sequences, already performed, have been reported in this paper. The first results are nevertheless very interesting, both from a scientific and nuclear utility point of view

  7. A High Precision Prediction Model Using Hybrid Grey Dynamic Model

    Li, Guo-Dong; Yamaguchi, Daisuke; Nagai, Masatake; Masuda, Shiro

    2008-01-01

    In this paper, we propose a new prediction analysis model which combines the first order one variable Grey differential equation Model (abbreviated as GM(1,1) model) from grey system theory and time series Autoregressive Integrated Moving Average (ARIMA) model from statistics theory. We abbreviate the combined GM(1,1) ARIMA model as ARGM(1,1)…

  8. Modelling trends in road accident frequency - Bayesian inference for rates with uncertain exposure

    Lloyd, Louise

    2013-01-01

    Several thousand people die as a result of a road accident each year in Great Britain and the trend in the number of fatal accidents is monitored closely to understand increases and reductions in the number of deaths. Results from analysis of these data directly influence Government road safety policy and ensure theintroduction of effective safety interventions across the country. Overall accident numbers are important, but when disaggregating into various characteristics, accident risk (def...

  9. Caries risk assessment models in caries prediction

    Amila Zukanović

    2013-01-01

    Objective. The aim of this research was to assess the efficiency of different multifactor models in caries prediction. Material and methods. Data from the questionnaire and objective examination of 109 examinees was entered into the Cariogram, Previser and Caries-Risk Assessment Tool (CAT) multifactor risk assessment models. Caries risk was assessed with the help of all three models for each patient, classifying them as low, medium or high-risk patients. The development of new caries lesions ...

  10. Current predictions for oil spill models

    Development and application of a background field of surface currents and a wind response model for oil spill software programs to predict the motion of an oil spill is described. The model determines the surface, seasonal and baroclinic currents. It uses input from all observed profiles of ocean density data for (in this case) the British Columbia coast. An objective analysis routine is used to prepare the spatially continuous, gridded fields of temperature and salinity from surface to ocean bottom. The model is evaluated by interpolating the wind field from weather buoy observations made in 1991, and a field of surface currents computed from tracks of Loran-C drifters deployed at the same time. Although the combined least squares fit does not fully explain the current variance, it does provide useful prediction based on parameters that can be embedded in search and rescue and oil spill prediction software. 14 refs., 2 tabs., 12 figs

  11. Modeling Advanced Neutron Source reactor station blackout accident using RELAP5

    The Advanced Neutron Source (ANS) system model using RELAP5 has been developed to perform loss-of-coolant accident (LOCA) and non-LOCA transients as safety-related input for early design considerations. The transients studies include LOCA, station blackout, and reactivity insertion accidents. The small-, medium-, and large-break LOCA results were presented and documented. This paper will focus on the station blackout scenario. The station blackout analyses have concentrated on thermal-hydraulic system response with and without accumulators. Five transient calculations were performed to characterize system performance using various numbers and sizes of accumulators at several key sites. The main findings will be discussed with recommendations for conceptual design considerations. ANS is a state-of-the-art research reactor to be built and operated at high heat flux, high mass flux, and high coolant subcooling. To accommodate these features, three ANS-specific changes were made in the RELAP5 code by adding: the Petukhov heat transfer correlation for single-phase forced convection in the thin coolant channel; the Gambill additive method with the Weatherhead wall superheat for the critical heat flux; and the Griffith drift flux model for the interfacial drag in the slug flow regime. 7 refs., 6 figs., 1 tab

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

    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)

  13. Mathematical modeling of ignition of woodlands resulted from accident on the pipeline

    Perminov, V. A.; Loboda, E. L.; Reyno, V. V.

    2014-11-01

    Accidents occurring at the sites of pipelines, accompanied by environmental damage, economic loss, and sometimes loss of life. In this paper we calculated the sizes of the possible ignition zones in emergency situations on pipelines located close to the forest, accompanied by the appearance of fireballs. In this paper, using the method of mathematical modeling calculates the maximum size of the ignition zones of vegetation as a result of accidental releases of flammable substances. The paper suggested in the context of the general mathematical model of forest fires give a new mathematical setting and method of numerical solution of a problem of a forest fire modeling. The boundary-value problem is solved numerically using the method of splitting according to physical processes. The dependences of the size of the forest fuel for different amounts of leaked flammable substances and moisture content of vegetation.

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

    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

  15. Multi-Model Ensemble Wake Vortex Prediction

    Koerner, Stephan; Holzaepfel, Frank; Ahmad, Nash'at N.

    2015-01-01

    Several multi-model ensemble methods are investigated for predicting wake vortex transport and decay. This study is a joint effort between National Aeronautics and Space Administration and Deutsches Zentrum fuer Luft- und Raumfahrt to develop a multi-model ensemble capability using their wake models. An overview of different multi-model ensemble methods and their feasibility for wake applications is presented. The methods include Reliability Ensemble Averaging, Bayesian Model Averaging, and Monte Carlo Simulations. The methodologies are evaluated using data from wake vortex field experiments.

  16. PREDICTIVE CAPACITY OF ARCH FAMILY MODELS

    Raphael Silveira Amaro

    2016-03-01

    Full Text Available In the last decades, a remarkable number of models, variants from the Autoregressive Conditional Heteroscedastic family, have been developed and empirically tested, making extremely complex the process of choosing a particular model. This research aim to compare the predictive capacity, using the Model Confidence Set procedure, than five conditional heteroskedasticity models, considering eight different statistical probability distributions. The financial series which were used refers to the log-return series of the Bovespa index and the Dow Jones Industrial Index in the period between 27 October 2008 and 30 December 2014. The empirical evidences showed that, in general, competing models have a great homogeneity to make predictions, either for a stock market of a developed country or for a stock market of a developing country. An equivalent result can be inferred for the statistical probability distributions that were used.

  17. A Predictive Model for Root Caries Incidence.

    Ritter, André V; Preisser, John S; Puranik, Chaitanya P; Chung, Yunro; Bader, James D; Shugars, Daniel A; Makhija, Sonia; Vollmer, William M

    2016-01-01

    This study aimed to find the set of risk indicators best able to predict root caries (RC) incidence in caries-active adults utilizing data from the Xylitol for Adult Caries Trial (X-ACT). Five logistic regression models were compared with respect to their predictive performance for incident RC using data from placebo-control participants with exposed root surfaces at baseline and from two study centers with ancillary data collection (n = 155). Prediction performance was assessed from baseline variables and after including ancillary variables [smoking, diet, use of removable partial dentures (RPD), toothbrush use, income, education, and dental insurance]. A sensitivity analysis added treatment to the models for both the control and treatment participants (n = 301) to predict RC for the control participants. Forty-nine percent of the control participants had incident RC. The model including the number of follow-up years at risk, the number of root surfaces at risk, RC index, gender, race, age, and smoking resulted in the best prediction performance, having the highest AUC and lowest Brier score. The sensitivity analysis supported the primary analysis and gave slightly better performance summary measures. The set of risk indicators best able to predict RC incidence included an increased number of root surfaces at risk and increased RC index at baseline, followed by white race and nonsmoking, which were strong nonsignificant predictors. Gender, age, and increased number of follow-up years at risk, while included in the model, were also not statistically significant. The inclusion of health, diet, RPD use, toothbrush use, income, education, and dental insurance variables did not improve the prediction performance. PMID:27160516

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

    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

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

    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

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

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

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

    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)

  2. Predictive coding as a model of cognition.

    Spratling, M W

    2016-08-01

    Previous work has shown that predictive coding can provide a detailed explanation of a very wide range of low-level perceptual processes. It is also widely believed that predictive coding can account for high-level, cognitive, abilities. This article provides support for this view by showing that predictive coding can simulate phenomena such as categorisation, the influence of abstract knowledge on perception, recall and reasoning about conceptual knowledge, context-dependent behavioural control, and naive physics. The particular implementation of predictive coding used here (PC/BC-DIM) has previously been used to simulate low-level perceptual behaviour and the neural mechanisms that underlie them. This algorithm thus provides a single framework for modelling both perceptual and cognitive brain function. PMID:27118562

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

    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

  4. Modelling the predictive performance of credit scoring

    Shi-Wei Shen

    2013-02-01

    Full Text Available Orientation: The article discussed the importance of rigour in credit risk assessment.Research purpose: The purpose of this empirical paper was to examine the predictive performance of credit scoring systems in Taiwan.Motivation for the study: Corporate lending remains a major business line for financial institutions. However, in light of the recent global financial crises, it has become extremely important for financial institutions to implement rigorous means of assessing clients seeking access to credit facilities.Research design, approach and method: Using a data sample of 10 349 observations drawn between 1992 and 2010, logistic regression models were utilised to examine the predictive performance of credit scoring systems.Main findings: A test of Goodness of fit demonstrated that credit scoring models that incorporated the Taiwan Corporate Credit Risk Index (TCRI, micro- and also macroeconomic variables possessed greater predictive power. This suggests that macroeconomic variables do have explanatory power for default credit risk.Practical/managerial implications: The originality in the study was that three models were developed to predict corporate firms’ defaults based on different microeconomic and macroeconomic factors such as the TCRI, asset growth rates, stock index and gross domestic product.Contribution/value-add: The study utilises different goodness of fits and receiver operator characteristics during the examination of the robustness of the predictive power of these factors.

  5. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert's law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions

  6. Preliminary design report for modeling of hydrogen uptake in fuel rod cladding during severe accidents

    Siefken, L.J.

    1998-08-01

    Preliminary designs are described for models of the interaction of Zircaloy and hydrogen and the consequences of this interaction on the behavior of fuel rod cladding during severe accidents. The modeling of this interaction and its consequences involves the modeling of seven processes: (1) diffusion of oxygen from the bulk gas into the boundary layer at the external cladding surface, (2) diffusion from the boundary layer into the oxide layer at the cladding external surface, (3) diffusion from the inner surface of the oxide layer into the metallic part of the cladding, (4) uptake of hydrogen in the event that the cladding oxide layer is dissolved in a steam-starved region, (5) embrittlement of cladding due to hydrogen uptake, (6) cracking of cladding during quenching due to its embrittlement and (7) release of hydrogen from the cladding after cracking of the cladding. An integral diffusion method is described for calculating the diffusion processes in the cladding. Experimental and theoretical results are presented that show the uptake of hydrogen in the event of dissolution of the oxide layer occurs rapidly and that show the release of hydrogen in the event of cracking of the cladding occurs rapidly. These experimental results are used as a basis for calculating the rate of hydrogen uptake and the rate of hydrogen release. The uptake of hydrogen is limited to the equilibrium solubility calculated by applying Sievert`s law. The uptake of hydrogen is an exothermic reaction that accelerates the heatup of a fuel rod. An embrittlement criteria is described that accounts for hydrogen and oxygen concentration and the extent of oxidation. A design is described for implementing the models for Zr-H interaction into the programming framework of the SCDAP/RELAP5 code. A test matrix is described for assessing the impact of the Zr-H interaction models on the calculated behavior of fuel rods in severe accident conditions.

  7. Modelling language evolution: Examples and predictions

    Gong, Tao; Shuai, Lan; Zhang, Menghan

    2014-06-01

    We survey recent computer modelling research of language evolution, focusing on a rule-based model simulating the lexicon-syntax coevolution and an equation-based model quantifying the language competition dynamics. We discuss four predictions of these models: (a) correlation between domain-general abilities (e.g. sequential learning) and language-specific mechanisms (e.g. word order processing); (b) coevolution of language and relevant competences (e.g. joint attention); (c) effects of cultural transmission and social structure on linguistic understandability; and (d) commonalities between linguistic, biological, and physical phenomena. All these contribute significantly to our understanding of the evolutions of language structures, individual learning mechanisms, and relevant biological and socio-cultural factors. We conclude the survey by highlighting three future directions of modelling studies of language evolution: (a) adopting experimental approaches for model evaluation; (b) consolidating empirical foundations of models; and (c) multi-disciplinary collaboration among modelling, linguistics, and other relevant disciplines.

  8. Global Solar Dynamo Models: Simulations and Predictions

    Mausumi Dikpati; Peter A. Gilman

    2008-03-01

    Flux-transport type solar dynamos have achieved considerable success in correctly simulating many solar cycle features, and are now being used for prediction of solar cycle timing and amplitude.We first define flux-transport dynamos and demonstrate how they work. The essential added ingredient in this class of models is meridional circulation, which governs the dynamo period and also plays a crucial role in determining the Sun’s memory about its past magnetic fields.We show that flux-transport dynamo models can explain many key features of solar cycles. Then we show that a predictive tool can be built from this class of dynamo that can be used to predict mean solar cycle features by assimilating magnetic field data from previous cycles.

  9. Grey Model for Stream Flow Prediction

    P. Syamala

    2012-04-01

    Full Text Available Design, operation and planning of water resources, irrigation and water supply systems require estimation of stream flow. A grey system or stochastic approach is required for dealing with the hydrological complexities of mid and long-term stream flow prediction. Generally relatively long period data series of stream flow records is required for the prediction using stochastic methods. In developing countries like India, availability of long period hydrological records is a problem. Grey system theory is applicable in the case of unclear innerrelationship, uncertain mechanisms and insufficient information and requires only small samples for parameter estimation. Stream flow records of Bharathapuzha river basin, Kerala, India is subjected to grey analysis. Model parameters were estimated using least-squares method. Statistical indices for the developed models indicate their ability to predict stream flow in the river under study with reasonable accuracy

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

    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.)

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

    Jung, Gunhyo; Kim, Juyoul [Seoul National University, Seoul (Korea, Republic of); Shin, Hyeongki [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

    2007-10-15

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

  12. An exponential filter model predicts lightness illusions

    Astrid eZeman

    2015-06-01

    Full Text Available Lightness, or perceived reflectance of a surface, is influenced by surrounding context. This is demonstrated by the Simultaneous Contrast Illusion (SCI, where a grey patch is perceived lighter against a black background and vice versa. Conversely, assimilation is where the lightness of the target patch moves towards that of the bounding areas and can be demonstrated in White's effect. Blakeslee and McCourt (2007 introduced an oriented difference-of-Gaussian (ODOG model that is able to account for both contrast and assimilation in a number of lightness illusions and that has been subsequently improved using localized normalization techniques. We introduce a model inspired by image statistics that is based on a family of exponential filters, with kernels spanning across multiple sizes and shapes. We include an optional second stage of normalization based on contrast gain control. Our model was tested on a well-known set of lightness illusions that have previously been used to evaluate ODOG and its variants, and model lightness values were compared with typical human data. We investigate whether predictive success depends on filters of a particular size or shape and whether pooling information across filters can improve performance. The best single filter correctly predicted the direction of lightness effects for 21 out of 27 illusions. Combining two filters together increased the best performance to 23, with asymptotic performance at 24 for an arbitrarily large combination of filter outputs. While normalization improved prediction magnitudes, it only slightly improved overall scores in direction predictions. The prediction performance of 24 out of 27 illusions equals that of the best performing ODOG variant, with greater parsimony. Our model shows that V1-style orientation-selectivity is not necessary to account for lightness illusions and that a low-level model based on image statistics is able to account for a wide range of both contrast and

  13. Model Predictive Control of a Tricopter

    Barsk, Karl-Johan

    2012-01-01

    In this master thesis, a real-time control system that stabilizes the rotational rates of a tri-copter, has been studied. The tricopter is a rotorcraft with three rotors. The tricopter has been modelled and identified, using system identification algorithms. The model has been used in a Kalman filter to estimate the state of the system and for design ofa model based controller. The control approach used in this thesis is a model predictive controller, which is a multi-variable controller that...

  14. Simulation error models for improved reservoir prediction

    Successful reservoir prediction requires an accurate estimation of parameters to be used in the reservoir model. This research focuses on developing models for simulation error within the petroleum industry, enabling accurate parameter estimation. The standard approach in the oil industry to parameter estimation in a Bayesian framework includes inappropriate assumptions about the error data. This leads to the parameter estimations being biased and overconfident. An error model is designed to significantly reduce the bias effect and to estimate an accurate range of spread. A 2D viscous fingering example problem will be used to demonstrate both construction of the error model, and the benefits gained in doing so

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

    Kim, Minsik; Ohba, Ryohji; Oura, Masamichi;

    2015-01-01

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

  16. A model for non-volatile fission product release during reactor accident conditions

    An analytical model has been developed to describe the release kinetics of non-volatile fission products (e.g., Mo, Ce, Ru and Ba) from uranium dioxide fuel under severe reactor accident conditions. The present treatment considers the rate-controlling process of release in accordance with diffusional transport in the fuel matrix and fission product vaporization from the fuel surface into the surrounding gas atmosphere. The effect of the oxygen potential in the gas atmosphere on the chemical form and volatility of the fission product is considered. A correlation is also developed to account for the trapping effects of Sb and Te in the Zircaloy cladding. This model has been used to interpret the release behaviour of fission products observed in the CEA experiments conducted in the HEVA/VERCORS facility at high temperature in a hydrogen and steam atmosphere. (author)

  17. modeling of a total loss pool water accident in mtr reactor

    in this study , it is intended to analyze early phases of a protected loss of coolant accident (LOCA)for MTR reactor. and to show the applicability of the presented model to the other similar types of research reactors. the transient situation since the time when coolant is beginning to be lost throughout one or more of the main coolant pipes which were supposed to be broken guillotine-like to the time when the core is totally uncovered is investigated. the modeling of the problem was separated into two phases; in the first phase when the water level of the pool is being decreased in a pre-estimated time -dependent way calculated by using modified Bernoulli equation, the conservation equation are solved by using shooting method. the later phase, when water level reaches the top level of fuel plates and begins to decrease until bottom of the core, and the fuel plates are being cooled by air.

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

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

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

    In a nuclear power plant, much knowledge on severe accidents has been acquired through PSA, and accident management (AM) guidelines are prepared by incorporating that knowledge. In PSA, it is necessary to evaluate the effectiveness of AM using the decision-making failure probability (DFP) of an emergency organization, operation failure probability of operators, success criteria of AM and reliability of AM equipment. However, to date there has been no suitable quantification method for PSA to obtain DFP. In this study, we developed a new method for DFP quantification of an emergency organization using a cognitive analysis model, and tried to apply it to S2DC and TMLF sequence of a typical plant. As a result: (1) The methods enabled to DFP quantification appropriate to level 1.5PSA by choosing the suitable value of a basic failure probability and an error factor. (2) The DFPs of six AMs appeared to be in the range of 0.23 to 0.41 (screening method) and in the range of 0.10 to 0.19 (detailed method), and the DFP decreased about 50% as a result of sensitivity analysis of the conservative assumption. (3) The screening method was more conservative than the detailed method, and it was shown to satisfy the screening performance required by PSA. (author)

  20. DKIST Polarization Modeling and Performance Predictions

    Harrington, David

    2016-05-01

    Calibrating the Mueller matrices of large aperture telescopes and associated coude instrumentation requires astronomical sources and several modeling assumptions to predict the behavior of the system polarization with field of view, altitude, azimuth and wavelength. The Daniel K Inouye Solar Telescope (DKIST) polarimetric instrumentation requires very high accuracy calibration of a complex coude path with an off-axis f/2 primary mirror, time dependent optical configurations and substantial field of view. Polarization predictions across a diversity of optical configurations, tracking scenarios, slit geometries and vendor coating formulations are critical to both construction and contined operations efforts. Recent daytime sky based polarization calibrations of the 4m AEOS telescope and HiVIS spectropolarimeter on Haleakala have provided system Mueller matrices over full telescope articulation for a 15-reflection coude system. AEOS and HiVIS are a DKIST analog with a many-fold coude optical feed and similar mirror coatings creating 100% polarization cross-talk with altitude, azimuth and wavelength. Polarization modeling predictions using Zemax have successfully matched the altitude-azimuth-wavelength dependence on HiVIS with the few percent amplitude limitations of several instrument artifacts. Polarization predictions for coude beam paths depend greatly on modeling the angle-of-incidence dependences in powered optics and the mirror coating formulations. A 6 month HiVIS daytime sky calibration plan has been analyzed for accuracy under a wide range of sky conditions and data analysis algorithms. Predictions of polarimetric performance for the DKIST first-light instrumentation suite have been created under a range of configurations. These new modeling tools and polarization predictions have substantial impact for the design, fabrication and calibration process in the presence of manufacturing issues, science use-case requirements and ultimate system calibration

  1. Predictive Modeling of the CDRA 4BMS

    Coker, Robert; Knox, James

    2016-01-01

    Fully predictive models of the Four Bed Molecular Sieve of the Carbon Dioxide Removal Assembly on the International Space Station are being developed. This virtual laboratory will be used to help reduce mass, power, and volume requirements for future missions. In this paper we describe current and planned modeling developments in the area of carbon dioxide removal to support future crewed Mars missions as well as the resolution of anomalies observed in the ISS CDRA.

  2. Linear Model Predictive Control of Induction Machine

    Mynář, Z.

    2015-01-01

    This article presents new control algorithm for induction machine based on linear model predictive control (MPC). Controller works in similar manners as field oriented control (FOC), but control is performed in stator coordinates. This reduces computational demands as Park’s transformation is absent and induction machine mathematical model in stator coordinates contains less nonlinear elements. Another aim of proposed controller was to achieve fast torque response.

  3. Model Predictive Control of Autonomous Vehicles

    Zanon, Mario; Frasch, Janick V.; Vukov, Milan; Sager, Sebastian; Diehl, Moritz

    2014-01-01

    International audience The control of autonomous vehicles is a challenging task that requires advanced control schemes. Nonlinear Model Predictive Control (NMPC) and Moving Horizon Estimation (MHE) are optimization-based control and estimation techniques that are able to deal with highly nonlinear, constrained, unstable and fast dynamic systems. In this chapter, these techniques are detailed, a descriptive nonlinear model is derived and the performance of the proposed control scheme is dem...

  4. Continuation model predictive control on smooth manifolds

    Knyazev, Andrew; Malyshev, Alexander,

    2015-01-01

    Model predictive control (MPC) anticipates future events to take appropriate control actions. Nonlinear MPC (NMPC) describes systems with nonlinear models and/or constraints. Continuation MPC, suggested by T.~Ohtsuka in 2004, uses Krylov-Newton iterations. Continuation MPC is suitable for nonlinear problems and has been recently adopted for minimum time problems. We extend the continuation MPC approach to a case where the state is implicitly constrained to a smooth manifold. We propose an alg...

  5. Preconditioning for continuation model predictive control

    Knyazev, Andrew; Malyshev, Alexander,

    2015-01-01

    Model predictive control (MPC) anticipates future events to take appropriate control actions. Nonlinear MPC (NMPC) deals with nonlinear models and/or constraints. A Continuation/GMRES Method for NMPC, suggested by T. Ohtsuka in 2004, uses the GMRES iterative algorithm to solve a forward difference approximation $Ax=b$ of the original NMPC equations on every time step. We have previously proposed accelerating the GMRES and MINRES convergence by preconditioning the coefficient matrix $A$. We no...

  6. Predictive performance models and multiple task performance

    Wickens, Christopher D.; Larish, Inge; Contorer, Aaron

    1989-01-01

    Five models that predict how performance of multiple tasks will interact in complex task scenarios are discussed. The models are shown in terms of the assumptions they make about human operator divided attention. The different assumptions about attention are then empirically validated in a multitask helicopter flight simulation. It is concluded from this simulation that the most important assumption relates to the coding of demand level of different component tasks.

  7. Cognitive modeling to predict video interpretability

    Young, Darrell L.; Bakir, Tariq

    2011-06-01

    Processing framework for cognitive modeling to predict video interpretability is discussed. Architecture consists of spatiotemporal video preprocessing, metric computation, metric normalization, pooling of like metric groups with masking adjustments, multinomial logistic pooling of Minkowski pooled groups of similar quality metrics, and estimation of confidence interval of final result.

  8. Predictive Modelling of Mycotoxins in Cereals

    Fels, van der H.J.; Liu, C.

    2015-01-01

    In dit artikel worden de samenvattingen van de presentaties tijdens de 30e bijeenkomst van de Werkgroep Fusarium weergegeven. De onderwerpen zijn: Predictive Modelling of Mycotoxins in Cereals.; Microbial degradation of DON.; Exposure to green leaf volatiles primes wheat against FHB but boosts produ

  9. Hierarchical Model Predictive Control for Resource Distribution

    Bendtsen, Jan Dimon; Trangbæk, K; Stoustrup, Jakob

    2010-01-01

    This paper deals with hierarchichal model predictive control (MPC) of distributed systems. A three level hierachical approach is proposed, consisting of a high level MPC controller, a second level of so-called aggregators, controlled by an online MPC-like algorithm, and a lower level of autonomous...

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

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

    2010-07-01

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

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

    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)

  12. Analyzing the causation of a railway accident based on a complex network

    In this paper, a new model is constructed for the causation analysis of railway accident based on the complex network theory. In the model, the nodes are defined as various manifest or latent accident causal factors. By employing the complex network theory, especially its statistical indicators, the railway accident as well as its key causations can be analyzed from the overall perspective. As a case, the “7.23” China—Yongwen railway accident is illustrated based on this model. The results show that the inspection of signals and the checking of line conditions before trains run played an important role in this railway accident. In conclusion, the constructed model gives a theoretical clue for railway accident prediction and, hence, greatly reduces the occurrence of railway accidents. (interdisciplinary physics and related areas of science and technology)

  13. Analyzing the causation of a railway accident based on a complex network

    Ma, Xin; Li, Ke-Ping; Luo, Zi-Yan; Zhou, Jin

    2014-02-01

    In this paper, a new model is constructed for the causation analysis of railway accident based on the complex network theory. In the model, the nodes are defined as various manifest or latent accident causal factors. By employing the complex network theory, especially its statistical indicators, the railway accident as well as its key causations can be analyzed from the overall perspective. As a case, the “7.23” China—Yongwen railway accident is illustrated based on this model. The results show that the inspection of signals and the checking of line conditions before trains run played an important role in this railway accident. In conclusion, the constructed model gives a theoretical clue for railway accident prediction and, hence, greatly reduces the occurrence of railway accidents.

  14. Radioiodine dosimetry and prediction of consequences of thyroid exposure of the Russian population following the Chernobyl accident

    In the early period after the Chernobyl accident, analysis of patterns of 131I exposure of the human thyroid showed that contaminated milk was the basic source of 131I intake among the inhabitants of Russia. The equipment and techniques used for measurement of the 131I content in the thyroids of these individuals are described in this work. A model of the 131I intake, taking into account protective actions, and a method of thyroid dose calculation are discussed. The mean thyroid dose and frequency distributions of the thyroid doses to inhabitants of towns and villages of the Bryansk, Tula and Orel regions of Russia are presented. The mean dose to the thyroids of children living in the villages was 2 to 5 times higher than the dose to adult thyroids; for children living in the towns, the mean dose was 1.5 to 12 times higher. The mean thyroid mass in adult inhabitants of the Bryansk region was 27 g, which exceeded the value for a standard man (20 g) and was taken into account in the dosimetric calculations. The technique for reconstructing the mean and individual thyroid doses was based on the correlation between thyroid dose and several parameters: Surface 137Cs activity in soil, dose rate in air in May of 1986, 131I content in local milk, milk consumption rate, and 134Cs + 137Cs content in the body. The collective thyroid dose to inhabitants of the most contaminated regions of Russia is estimated and a thyroid cancer rate prognosis is derived. The need for intensified medical care for the critical group - children of preschool age during 1986 - is based on a significant increase in the number of projected thyroid cancers and adenomas. 32 refs., 10 figs., 15 tabs

  15. Specialized Language Models using Dialogue Predictions

    Popovici, C; Popovici, Cosmin; Baggia, Paolo

    1996-01-01

    This paper analyses language modeling in spoken dialogue systems for accessing a database. The use of several language models obtained by exploiting dialogue predictions gives better results than the use of a single model for the whole dialogue interaction. For this reason several models have been created, each one for a specific system question, such as the request or the confirmation of a parameter. The use of dialogue-dependent language models increases the performance both at the recognition and at the understanding level, especially on answers to system requests. Moreover other methods to increase performance, like automatic clustering of vocabulary words or the use of better acoustic models during recognition, does not affect the improvements given by dialogue-dependent language models. The system used in our experiments is Dialogos, the Italian spoken dialogue system used for accessing railway timetable information over the telephone. The experiments were carried out on a large corpus of dialogues coll...

  16. Caries risk assessment models in caries prediction

    Amila Zukanović

    2013-11-01

    Full Text Available Objective. The aim of this research was to assess the efficiency of different multifactor models in caries prediction. Material and methods. Data from the questionnaire and objective examination of 109 examinees was entered into the Cariogram, Previser and Caries-Risk Assessment Tool (CAT multifactor risk assessment models. Caries risk was assessed with the help of all three models for each patient, classifying them as low, medium or high-risk patients. The development of new caries lesions over a period of three years [Decay Missing Filled Tooth (DMFT increment = difference between Decay Missing Filled Tooth Surface (DMFTS index at baseline and follow up], provided for examination of the predictive capacity concerning different multifactor models. Results. The data gathered showed that different multifactor risk assessment models give significantly different results (Friedman test: Chi square = 100.073, p=0.000. Cariogram is the model which identified the majority of examinees as medium risk patients (70%. The other two models were more radical in risk assessment, giving more unfavorable risk –profiles for patients. In only 12% of the patients did the three multifactor models assess the risk in the same way. Previser and CAT gave the same results in 63% of cases – the Wilcoxon test showed that there is no statistically significant difference in caries risk assessment between these two models (Z = -1.805, p=0.071. Conclusions. Evaluation of three different multifactor caries risk assessment models (Cariogram, PreViser and CAT showed that only the Cariogram can successfully predict new caries development in 12-year-old Bosnian children.

  17. Disease prediction models and operational readiness.

    Courtney D Corley

    Full Text Available The objective of this manuscript is to present a systematic review of biosurveillance models that operate on select agents and can forecast the occurrence of a disease event. We define a disease event to be a biological event with focus on the One Health paradigm. These events are characterized by evidence of infection and or disease condition. We reviewed models that attempted to predict a disease event, not merely its transmission dynamics and we considered models involving pathogens of concern as determined by the US National Select Agent Registry (as of June 2011. We searched commercial and government databases and harvested Google search results for eligible models, using terms and phrases provided by public health analysts relating to biosurveillance, remote sensing, risk assessments, spatial epidemiology, and ecological niche modeling. After removal of duplications and extraneous material, a core collection of 6,524 items was established, and these publications along with their abstracts are presented in a semantic wiki at http://BioCat.pnnl.gov. As a result, we systematically reviewed 44 papers, and the results are presented in this analysis. We identified 44 models, classified as one or more of the following: event prediction (4, spatial (26, ecological niche (28, diagnostic or clinical (6, spread or response (9, and reviews (3. The model parameters (e.g., etiology, climatic, spatial, cultural and data sources (e.g., remote sensing, non-governmental organizations, expert opinion, epidemiological were recorded and reviewed. A component of this review is the identification of verification and validation (V&V methods applied to each model, if any V&V method was reported. All models were classified as either having undergone Some Verification or Validation method, or No Verification or Validation. We close by outlining an initial set of operational readiness level guidelines for disease prediction models based upon established Technology

  18. An analysis of accidents involving towboat-barge combination on selected inland waterways of the United States.

    Gamble, William John

    1980-01-01

    Approved for public release; distribution is unlimited This study uses a statistical analysis approach on a computerized data base to analyze accidents involving towboat-barge combinations on the inland waterways of the United States. The main areas explored are the factors affecting the severity and the frequency of accidents. In addition, multiple regression models are used to predict the severity of towboat accidents from a set of independent accident variables. Conclusions and recom...

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

    Highlights: • MCCI research in SARNET2 consisted of large and small-scale testing and modelling. • Heavy (U,Zr)O2 melt interacts with light concrete oxide to produce multiple, mixed phases. • Anisotropic concrete erosion of high SiO2 concrete by melts may be linked with crust stability. • Understanding this anisotropy is important for reliable 3-D modelling. • Precise measures of refractory oxide data greatly improve phase diagram accuracy. - Abstract: In a severe accident transient, corium composition and its properties determine its behaviour and its potential interactions both with the reactor vessel and in the later phases with the concrete basemat. This, in turn, requires a detailed knowledge of the phases present at temperature and how they are formed. Because it implies mainly the investigation of chemical systems at high temperature, these data are often difficult to obtain or are uncertain if it already exists. Therefore more data are required both to complete the thermodynamic databanks (such as NUCLEA) and to construct accurate equilibrium phase diagrams and to finally contribute to the improvement of the codes simulating these severe accident conditions. The MCCI work package (WP6) of the SARNET 2 Network of Excellence has been addressing these problems. In this framework in large facilities such as VULCANO tests have been performed on the interactions and ablation of UO2-containing melts with concrete. They have been completed by large scale MCCI testing such EPICOR on vessel steel corrosion. In parallel in major EU-funded ISTC projects co-ordinated with national institutes, such as the CORPHAD and PRECOS, smaller, single effect tests have been carried out on the more difficult phase diagrams. These have produced data that can be directly used by databanks and for modelling improvement/validation. From these data significant advances in the melt chemistry and pool behaviour have been made. A selection of experiments from participating

  20. Pretest analysis of containment studies facility model for simulated loss of coolant accident conditions

    An experimental facility called Containment Studies Facility (CSF) has been constructed at Bhabha Atomic Research Centre (BARC), Trombay for the purpose of research and development in the area of nuclear reactor containment thermal hydraulics. The facility consists of reinforced concrete containment structural model and a Primary Heat Transport Model (PHTM) vessel. The containment model is approximately 1:250 volumetrically scaled down model of a 220 MWe Indian Pressurized Heavy Water Reactor (IPHWR) containment system and the PHTM represents the primary heat transport system of the prototype reactor. The PHTM with a pressure vessel and associated pump and piping system is designed for simulating the Loss of Coolant Accident (LOCA) or Main Steam Line Break (MSLB) conditions within the containment model. As part of CSF project thermal hydraulic analysis, a pretest analysis was carried out for simulated LOCA conditions. Blow down mass and energy discharge data were obtained using Relap/MOD3.2 code for different blow down conditions and were used as inputs to CONTRAN code for simulating LOCA or main steam line break (MSLB) conditions in the containment model. Pressure and temperature transients in the CSF model for different blow down conditions and a number of parametric studies were conducted to assess the influence of a large number of thermodynamic and geometrical parameters which are known to affect the transients and alter the peak pressure and temperature values. (author)

  1. Model Predictive Control based on Finite Impulse Response Models

    Prasath, Guru; Jørgensen, John Bagterp

    2008-01-01

    We develop a regularized l2 finite impulse response (FIR) predictive controller with input and input-rate constraints. Feedback is based on a simple constant output disturbance filter. The performance of the predictive controller in the face of plant-model mismatch is investigated by simulations...

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

    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 corr...... method is proposed for the long-range radiological consequences of nuclear accidents. (C) 1999 Elsevier Science Ltd. All rights reserved.......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...

  3. STELLA Experiment: Design and Model Predictions

    The STaged ELectron Laser Acceleration (STELLA) experiment will be one of the first to examine the critical issue of staging the laser acceleration process. The BNL inverse free electron laser (EEL) will serve as a prebuncher to generate ∼ 1 (micro)m long microbunches. These microbunches will be accelerated by an inverse Cerenkov acceleration (ICA) stage. A comprehensive model of the STELLA experiment is described. This model includes the EEL prebunching, drift and focusing of the microbunches into the ICA stage, and their subsequent acceleration. The model predictions will be presented including the results of a system error study to determine the sensitivity to uncertainties in various system parameters

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

    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.

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

    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

  6. Applying hierarchical loglinear models to nonfatal underground coal mine accidents for safety management.

    Onder, Mustafa; Onder, Seyhan; Adiguzel, Erhan

    2014-01-01

    Underground mining is considered to be one of the most dangerous industries and mining remains the most hazardous occupation. Categorical analysis of accident records may present valuable information for preventing accidents. In this study, hierarchical loglinear analysis was applied to occupational injuries that occurred in an underground coal mine. The main factors affecting the accidents were defined as occupation, area, reason, accident time and part of body affected. By considering subfactors of the main factors, multiway contingency tables were prepared and, thus, the probabilities that might affect nonfatal injuries were investigated. At the end of the study, important accident risk factors and job groups with a high probability of being exposed to those risk factors were determined. This article presents important information on decreasing the number accidents in underground coal mines. PMID:24934420

  7. Performance model to predict overall defect density

    J Venkatesh

    2012-08-01

    Full Text Available Management by metrics is the expectation from the IT service providers to stay as a differentiator. Given a project, the associated parameters and dynamics, the behaviour and outcome need to be predicted. There is lot of focus on the end state and in minimizing defect leakage as much as possible. In most of the cases, the actions taken are re-active. It is too late in the life cycle. Root cause analysis and corrective actions can be implemented only to the benefit of the next project. The focus has to shift left, towards the execution phase than waiting for lessons to be learnt post the implementation. How do we pro-actively predict defect metrics and have a preventive action plan in place. This paper illustrates the process performance model to predict overall defect density based on data from projects in an organization.

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

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

    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.

  10. Predicting the long-term 137Cs distribution in Fukushima after the Fukushima Dai-ichi nuclear power plant accident: a parameter sensitivity analysis

    than those of the other rivers. Annual sediment outflows from the Abukuma River and the total from the other 13 river basins were calculated as 3.2 × 104–3.1 × 105 and 3.4 × 104–2.1 × 105 t y−1, respectively. The values vary between calculation cases because of the critical shear stress, the rainfall factor, and other differences. On the other hand, contributions of those parameters were relatively small for 137Cs concentration within transported soil. This indicates that the total amount of 137Cs outflow into the ocean would mainly be controlled by the amount of soil erosion and transport and the total amount of 137Cs concentration remaining within the basin. Outflows of 137Cs from the Abukuma River and the total from the other 13 river basins during the first year after the accident were calculated to be 2.3 × 1011–3.7 × 1012 and 4.6 × 1011–6.5 × 1012 Bq y−1, respectively. The former results were compared with the field investigation results, and the order of magnitude was matched between the two, but the value of the investigation result was beyond the upper limit of model prediction. - Highlights: • A simulation model to predict future distribution of contaminants is developed. • Model utilizes Geographical Information System (GIS) to integrate online open data. • Model simulates soil erosion, transport and sedimentation on surface of Fukushima. • Simulation results were found to be qualitatively consistent with existing data

  11. Neuro-fuzzy modeling in bankruptcy prediction

    Vlachos D.

    2003-01-01

    Full Text Available For the past 30 years the problem of bankruptcy prediction had been thoroughly studied. From the paper of Altman in 1968 to the recent papers in the '90s, the progress of prediction accuracy was not satisfactory. This paper investigates an alternative modeling of the system (firm, combining neural networks and fuzzy controllers, i.e. using neuro-fuzzy models. Classical modeling is based on mathematical models that describe the behavior of the firm under consideration. The main idea of fuzzy control, on the other hand, is to build a model of a human control expert who is capable of controlling the process without thinking in a mathematical model. This control expert specifies his control action in the form of linguistic rules. These control rules are translated into the framework of fuzzy set theory providing a calculus, which can stimulate the behavior of the control expert and enhance its performance. The accuracy of the model is studied using datasets from previous research papers.

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

    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

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

    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. Progresses in tritium accident modelling in the frame of IAEA EMRAS II

    The assessment of the environmental impact of tritium release from nuclear facilities is a topic of interest in many countries. In the IAEA's Environmental Modelling for Radiation Safety (EMRAS I) programme, progresses for routine releases were done and in the EMRAS II programme a dedicated working group (WG 7 - Tritium Accidents) focused on the potential accidental releases (liquid and atmospheric pathways). The progresses achieved in WG 7 were included in a complex report - a technical document of IAEA covering both liquid and atmospheric accidental release consequences. A brief description of the progresses achieved in the frame of EMRAS II WG 7 is presented. Important results have been obtained concerning washout rate, the deposition on the soil of HTO and HT, the HTO uptake by leaves and the subsequent conversion to OBT (organically bound tritium) during daylight. Further needs of the processes understanding and the experimental efforts are emphasised

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

    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.

  16. Prediction of vascular cerebral accidents by PET T.D.M. with {sup 18}F-F.D.G; Prediction des accidents vasculaires cerebraux par la TEP -TDM vasculaire au 18F-FDG

    Grandpierre, S.; Chevalier, O.; Thomas, V.; Netter, F.; Meneroux, B.; Karcher, G.; Marie, P.Y. [Service de medecine nucleaire, CHU de Nancy, (France); Desandes, E. [departement d' informatique medical, centre Alexis-Vautrin, Nancy, (France)

    2009-05-15

    This study is the first to show a relationship between the vascular captation of the F.D.G. in PET and the risk of a later ischemic cerebral vascular accident. this relation seems particularly strong for the sources of the carotids junction, so that the PET with F.D.G. could be useful to evaluate the stability of atheromas injuries in this area. (N.C.)

  17. Monitoring Severe Accidents Using AI Techniques

    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. Disease Prediction Models and Operational Readiness

    Corley, Courtney D.; Pullum, Laura L.; Hartley, David M.; Benedum, Corey M.; Noonan, Christine F.; Rabinowitz, Peter M.; Lancaster, Mary J.

    2014-03-19

    INTRODUCTION: The objective of this manuscript is to present a systematic review of biosurveillance models that operate on select agents and can forecast the occurrence of a disease event. One of the primary goals of this research was to characterize the viability of biosurveillance models to provide operationally relevant information for decision makers to identify areas for future research. Two critical characteristics differentiate this work from other infectious disease modeling reviews. First, we reviewed models that attempted to predict the disease event, not merely its transmission dynamics. Second, we considered models involving pathogens of concern as determined by the US National Select Agent Registry (as of June 2011). Methods: We searched dozens of commercial and government databases and harvested Google search results for eligible models utilizing terms and phrases provided by public health analysts relating to biosurveillance, remote sensing, risk assessments, spatial epidemiology, and ecological niche-modeling, The publication date of search results returned are bound by the dates of coverage of each database and the date in which the search was performed, however all searching was completed by December 31, 2010. This returned 13,767 webpages and 12,152 citations. After de-duplication and removal of extraneous material, a core collection of 6,503 items was established and these publications along with their abstracts are presented in a semantic wiki at http://BioCat.pnnl.gov. Next, PNNL’s IN-SPIRE visual analytics software was used to cross-correlate these publications with the definition for a biosurveillance model resulting in the selection of 54 documents that matched the criteria resulting Ten of these documents, However, dealt purely with disease spread models, inactivation of bacteria, or the modeling of human immune system responses to pathogens rather than predicting disease events. As a result, we systematically reviewed 44 papers and the

  19. Structural evaluation of electrosleeved tubes under severe accident transients

    A flow stress model was developed for predicting failure of Electrosleeved PWR steam generator tubing under severe accident transients. The Electrosleeve, which is nanocrystalline pure nickel, loses its strength at temperatures greater than 400 C during severe accidents because of grain growth. A grain growth model and the Hall-Petch relationship were used to calculate the loss of flow stress as a function of time and temperature during the accident. Available tensile test data as well as high temperature failure tests on notched Electrosleeved tube specimens were used to derive the basic parameters of the failure model. The model was used to predict the failure temperatures of Electrosleeved tubes with axial cracks in the parent tube during postulated severe accident transients

  20. Developing Predictive Models of Health Literacy

    Martin, Laurie T.; Ruder, Teague; Escarce, José J.; Ghosh-Dastidar, Bonnie; Sherman, Daniel; Elliott, Marc; Bird, Chloe E.; Fremont, Allen; Gasper, Charles; Culbert, Arthur; Lurie, Nicole

    2009-01-01

    INTRODUCTION Low health literacy (LHL) remains a formidable barrier to improving health care quality and outcomes. Given the lack of precision of single demographic characteristics to predict health literacy, and the administrative burden and inability of existing health literacy measures to estimate health literacy at a population level, LHL is largely unaddressed in public health and clinical practice. To help overcome these limitations, we developed two models to estimate health literacy. ...

  1. Modelling molecular flexibility for crystal structure prediction

    Uzoh, O. G.

    2015-01-01

    In the crystal packing of molecules wherein a single bond links aromatic groups, a change in the torsion angle can optimise close packing of the molecule. The improved intermolecular interactions, Uinter, outweigh the conformational energy penalty, ΔEintra, to give a more stable lattice energy, Elatt = Uinter + ΔEintra. This thesis uses this lattice energy model hierarchically in a new Crystal Structure Prediction (CSP) algorithm, CrystalPredictor version 1.6, which varies the low-barrier tor...

  2. Nonlinear model predictive control using automatic differentiation

    Al Seyab, Rihab Khalid Shakir

    2006-01-01

    Although nonlinear model predictive control (NMPC) might be the best choice for a nonlinear plant, it is still not widely used. This is mainly due to the computational burden associated with solving online a set of nonlinear differential equations and a nonlinear dynamic optimization problem in real time. This thesis is concerned with strategies aimed at reducing the computational burden involved in different stages of the NMPC such as optimization problem, state estimation, an...

  3. Improvement in post test accident analysis results prediction for the test no. 2 in PSB test facility by applying UMAE methodology

    This paper mainly deals with the improvement in the post test accident analysis results prediction for the test no. 2, 'Total loss of feed water with failure of HPIS pumps and operator actions on primary and secondary circuit depressurization', carried-out on PSB integral test facility in May 2005. This is one the most complicated test conducted in PSB test facility. The prime objective of this test is to provide support for the verification of the accident management strategies for NPPs and also to verify the correctness of some safety systems operating only during accident. The objective of this analysis is to assess the capability to reproduce the phenomena occurring during the selected tests and to quantify the accuracy of the code calculation qualitatively and quantitatively for the best estimate code Relap5/mod3.3 by systematically applying all the procedures lead by Uncertainty Methodology based on Accuracy Extrapolation (UMAE), developed at University of Pisa. In order to achieve these objectives test facility nodalisation qualification for both 'steady state level' and 'on transient level' are demonstrated. For the 'steady state level' qualification compliance to acceptance criteria established in UMAE has been checked for geometrical details and thermal hydraulic parameters. The following steps have been performed for evaluation of qualitative qualification of 'on transient level': visual comparisons between experimental and calculated relevant parameters time trends; list of comparison between experimental and code calculation resulting time sequence of significant events; identification/verification of CSNI phenomena validation matrix; use of the Phenomenological Windows (PhW), identification of Key Phenomena and Relevant Thermal-hydraulic Aspects (RTA). A successful application of the qualitative process constitutes a prerequisite to the application of the quantitative analysis. For quantitative accuracy of code prediction Fast Fourier Transform Based

  4. Predicting extinction rates in stochastic epidemic models

    We investigate the stochastic extinction processes in a class of epidemic models. Motivated by the process of natural disease extinction in epidemics, we examine the rate of extinction as a function of disease spread. We show that the effective entropic barrier for extinction in a susceptible–infected–susceptible epidemic model displays scaling with the distance to the bifurcation point, with an unusual critical exponent. We make a direct comparison between predictions and numerical simulations. We also consider the effect of non-Gaussian vaccine schedules, and show numerically how the extinction process may be enhanced when the vaccine schedules are Poisson distributed

  5. Thermohydrodynamic models adequacy assessment methods within the frameworks of a calculation means verification/validation program for accident processes analysis

    Within the frameworks of the previous developed by authors generalised calculation means (codes) verification / validation methodology in given article considers procedure realisation that related to mathematical thermohydrodynamic models adequacy analysis to real processes of accident / transition modes.For concreteness the last RELAP5 modifications are considered as a calculation code,and VVER reactor equipment is considered as a study object

  6. Constructing predictive models of human running.

    Maus, Horst-Moritz; Revzen, Shai; Guckenheimer, John; Ludwig, Christian; Reger, Johann; Seyfarth, Andre

    2015-02-01

    Running is an essential mode of human locomotion, during which ballistic aerial phases alternate with phases when a single foot contacts the ground. The spring-loaded inverted pendulum (SLIP) provides a starting point for modelling running, and generates ground reaction forces that resemble those of the centre of mass (CoM) of a human runner. Here, we show that while SLIP reproduces within-step kinematics of the CoM in three dimensions, it fails to reproduce stability and predict future motions. We construct SLIP control models using data-driven Floquet analysis, and show how these models may be used to obtain predictive models of human running with six additional states comprising the position and velocity of the swing-leg ankle. Our methods are general, and may be applied to any rhythmic physical system. We provide an approach for identifying an event-driven linear controller that approximates an observed stabilization strategy, and for producing a reduced-state model which closely recovers the observed dynamics. PMID:25505131

  7. A model of the operator cognitive behaviors during the steam generator tube rupture accident at a nuclear power plant

    An integrated framework of modeling the human operator cognitive behavior during nuclear power plant accident scenarios is presented. It incorporates both plant and operator models. The basic structure of the operator model is similar to that of existing cognitive models, however, this model differs from those existing ones largely in two aspects. First, using frame and membership function, the pattern matching behavior, which is identified as the dominant cognitive process of operators responding to an accident sequence, is explicitly implemented in this model. Second, the non-task-related human cognitive activities like effects of stress and cognitive biases such as confirmation bias and availability bias, are also considered. A computer code, OPEC is assembled to simulate this framework and is actually applied to an SGTR sequence, and the resultant simulated behaviors of operator are obtained. 28 refs., 4 figs., 6 tabs. (author)

  8. Improved models for the simulation of severe LWR accidents - processes during quenching and chemical interactions

    In the Core Degradation Project, the contributions of the IKE mainly concerns the improvements and extensions of models, basic versions of which have been developed in the frame of the national BMBF - project KESS. In this project detailed models have been developed to simulate the main processes in the core during a severe accident in light water reactors. The first part of this report is focused on the interacting processes during quenching, like the embrittlement of the fuel rod cladding and of the refreezed melt, the oxidation of the cladding and the crust as well as the cooling effect due to the rapid vaporization. The improved and extended models have been implemented in the code system KESS and as a frist step of the validation the integral bundle experiment CORA-13 has been used. The second part of this report is directed to the chemical interaction between the fuel rod cladding and the Inconel grid spacer. Hereby, a basic diffusion model has been developed and applied to specific bundle conditions to take into account the time of failure of the grid spacer. (orig.)

  9. Study of heat and mass transfer phenomena in fuel assembly models under accident conditions

    The majority of the material in support of the thermal - hydraulic safety of WWER core was obtained on single - assembly models containing a relatively small number of elements - heater rods. Upgrading the requirements to the reactor safety leads to the necessity for studying phenomena in channels representing the cross - sectional core dimensions and non - uniform radial power generation. Under such conditions, the contribution of natural convection can be significant in some core zones, including the occurrence of reverse flows and interchannel instability. These phenomena can have an important influence on heat transfer processes. Such influence is especially drastical under accident conditions associated with ceasing the forced circulation over the circuit. A number of urgent reactor safety problems at low operating parameters is related with the computer code verification and certification. One of the important trends in the reactor safety research is concerned with the rod bundle reflooding and verificational calculations of this phenomenon. To assess the water cooled reactor safety, the best fit computer codes are employed, which make it possible to simulate accident and transient operating conditions in a reactor installation. One of the most widely known computer codes is the RELAP5/MOD3 Code. The paper presents the comparison of the results calculated using this computer code with the test data on 4 - rod bundle quenching, which were obtained at the SSCRF-IPPE. Recently, the investigations on the steam - zirconium reaction kinetics have been performed at the SSCFR-IPPE and are being presently performed for the purpose of developing new and verifying available computer codes. (author). 3 refs, 6 figs

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

    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

  11. Ship waste quantities prediction model for the port of Belgrade

    VLADANKA PRESBURGER ULNIKOVIĆ

    2011-06-01

    Full Text Available This study focuses on the issues related to the waste management in river ports in general and especially in the port of Belgrade. Data on solid waste, waste oils, oily waters, gray water and black water have been collected for a period of five years. The methodology of data collection is presented. Trends of data were analyzed and the regression model was used to predict the waste quantities in the Belgrade port. This data could be utilized as a basis for the calculation of the equipment capacity for waste selective collection, treatment and storage. The results presented in this study establish the need for an orga¬nized management system for this type of waste which can be achieved either by constructing and providing new specialized terminal or by providing mobile floating facilities and other plants in the Port of Belgrade for these kinds of ser¬vices. In addition to the above, the legislative and organizational strategy of waste management has been explored to complete the study because the im¬pact of good waste management on environment and prevention of environ¬mental accidents would be highly beneficial. This study demonstrated that ad¬dressing these issues should be considered at international as well as national level.

  12. 高速公路交通事故起数时空分析模型%Temporal-spatial analysis model of traffic accident frequency on expressway

    马壮林; 邵春福; 胡大伟; 马社强

    2012-01-01

    In order to analyze the relationships among traffic accident frequency and potential influencing factors such as time,road space structure and traffic running environment,nine independent variables were selected from the aspects of time and space,two kinds of section divided methods were adopted,which were fixed-length consistent segment and longitudinal grade consistent segment,and the hourly,weekly and monthly distribution models of traffic accident frequency were constructed.A typical accident-prone section was selected,and Poisson regression model,negative binomial regression model,zero-inflated Poisson regression model and zero-inflated negative binomial regression model were used to fit hourly,weekly and monthly distribution models respectively.The best forms of three models were determined,and the temporal-spatial analysis model of traffic accident frequency was established based on the goodness of fit test.Analysis result shows that the fitting effect of negative binomial regression model is better for traffic accident hourly and monthly distribution models based on fixed-length consistent segment from the views of AIC and BIC,and the fitting effect of Poisson regression model is better for other models.The prediction errors of traffic accident hourly,weekly and monthly distribution model based on fixed-length consistent segment are less than those of longitudinal grade consistent segment.4 tabs,15 refs.%为了分析交通事故起数与时间、道路空间结构及交通运行环境等潜在影响因素之间的关系,从时间和空间角度选择9个自变量,分别从路段长度一致和路段坡度一致2个角度,构建交通事故起数时段、周日和月分布模型。以某典型交通事故多发段为例,分别运用泊松回归模型、负二项回归模型、零堆积泊松回归模型和零堆积负二项回归模型拟合交通事故起数时段、周日和月分布模型,根据模型的拟合优度检验,分别确定3个模型的最佳

  13. Prediction of vascular cerebral accidents by PET T.D.M. with 18F-F.D.G

    This study is the first to show a relationship between the vascular captation of the F.D.G. in PET and the risk of a later ischemic cerebral vascular accident. this relation seems particularly strong for the sources of the carotids junction, so that the PET with F.D.G. could be useful to evaluate the stability of atheromas injuries in this area. (N.C.)

  14. Oil Spill Detection and Modelling: Preliminary Results for the Cercal Accident

    da Costa, R. T.; Azevedo, A.; da Silva, J. C. B.; Oliveira, A.

    2013-03-01

    Oil spill research has significantly increased mainly as a result of the severe consequences experienced from industry accidents. Oil spill models are currently able to simulate the processes that determine the fate of oil slicks, playing an important role in disaster prevention, control and mitigation, generating valuable information for decision makers and the population in general. On the other hand, satellite Synthetic Aperture Radar (SAR) imagery has demonstrated significant potential in accidental oil spill detection, when they are accurately differentiated from look-alikes. The combination of both tools can lead to breakthroughs, particularly in the development of Early Warning Systems (EWS). This paper presents a hindcast simulation of the oil slick resulting from the Motor Tanker (MT) Cercal oil spill, listed by the Portuguese Navy as one of the major oil spills in the Portuguese Atlantic Coast. The accident took place nearby Leix˜oes Harbour, North of the Douro River, Porto (Portugal) on the 2nd of October 1994. The oil slick was segmented from available European Remote Sensing (ERS) satellite SAR images, using an algorithm based on a simplified version of the K-means clustering formulation. The image-acquired information, added to the initial conditions and forcings, provided the necessary inputs for the oil spill model. Simulations were made considering the tri-dimensional hydrodynamics in a crossscale domain, from the interior of the Douro River Estuary to the open-ocean on the Iberian Atlantic shelf. Atmospheric forcings (from ECMWF - the European Centre for Medium-Range Weather Forecasts and NOAA - the National Oceanic and Atmospheric Administration), river forcings (from SNIRH - the Portuguese National Information System of the Hydric Resources) and tidal forcings (from LNEC - the National Laboratory for Civil Engineering), including baroclinic gradients (NOAA), were considered. The lack of data for validation purposes only allowed the use of the

  15. A predictive model for dimensional errors in fused deposition modeling

    Stolfi, A.

    2015-01-01

    values of L (0.254 mm, 0.330 mm) was produced by comparing predicted values with external face-to-face measurements. After removing outliers, the results show that the developed two-parameter model can serve as tool for modeling the FDM dimensional behavior in a wide range of deposition angles.......This work concerns the effect of deposition angle (a) and layer thickness (L) on the dimensional performance of FDM parts using a predictive model based on the geometrical description of the FDM filament profile. An experimental validation over the whole a range from 0° to 177° at 3° steps and two...

  16. Modeling and Prediction of Krueger Device Noise

    Guo, Yueping; Burley, Casey L.; Thomas, Russell H.

    2016-01-01

    This paper presents the development of a noise prediction model for aircraft Krueger flap devices that are considered as alternatives to leading edge slotted slats. The prediction model decomposes the total Krueger noise into four components, generated by the unsteady flows, respectively, in the cove under the pressure side surface of the Krueger, in the gap between the Krueger trailing edge and the main wing, around the brackets supporting the Krueger device, and around the cavity on the lower side of the main wing. For each noise component, the modeling follows a physics-based approach that aims at capturing the dominant noise-generating features in the flow and developing correlations between the noise and the flow parameters that control the noise generation processes. The far field noise is modeled using each of the four noise component's respective spectral functions, far field directivities, Mach number dependencies, component amplitudes, and other parametric trends. Preliminary validations are carried out by using small scale experimental data, and two applications are discussed; one for conventional aircraft and the other for advanced configurations. The former focuses on the parametric trends of Krueger noise on design parameters, while the latter reveals its importance in relation to other airframe noise components.

  17. Artificial Neural Network Model for Predicting Compressive

    Salim T. Yousif

    2013-05-01

    Full Text Available   Compressive strength of concrete is a commonly used criterion in evaluating concrete. Although testing of the compressive strength of concrete specimens is done routinely, it is performed on the 28th day after concrete placement. Therefore, strength estimation of concrete at early time is highly desirable. This study presents the effort in applying neural network-based system identification techniques to predict the compressive strength of concrete based on concrete mix proportions, maximum aggregate size (MAS, and slump of fresh concrete. Back-propagation neural networks model is successively developed, trained, and tested using actual data sets of concrete mix proportions gathered from literature.    The test of the model by un-used data within the range of input parameters shows that the maximum absolute error for model is about 20% and 88% of the output results has absolute errors less than 10%. The parametric study shows that water/cement ratio (w/c is the most significant factor  affecting the output of the model.     The results showed that neural networks has strong potential as a feasible tool for predicting compressive strength of concrete.

  18. Evaluating predictive models of software quality

    Applications from High Energy Physics scientific community are constantly growing and implemented by a large number of developers. This implies a strong churn on the code and an associated risk of faults, which is unavoidable as long as the software undergoes active evolution. However, the necessities of production systems run counter to this. Stability and predictability are of paramount importance; in addition, a short turn-around time for the defect discovery-correction-deployment cycle is required. A way to reconcile these opposite foci is to use a software quality model to obtain an approximation of the risk before releasing a program to only deliver software with a risk lower than an agreed threshold. In this article we evaluated two quality predictive models to identify the operational risk and the quality of some software products. We applied these models to the development history of several EMI packages with intent to discover the risk factor of each product and compare it with its real history. We attempted to determine if the models reasonably maps reality for the applications under evaluation, and finally we concluded suggesting directions for further studies.

  19. Effect of expansion wave on WWER type reactor model in loss-of-coolant accident

    An experimental device was developed for the investigation of the effect of the expansion wave arising during a loss-of-coolant accident on a WWER pressure vessel model (1:8). The device enables a maximum pressure of 12 MPa to be achieved in the system. Water heating is provided by electrical heating to 270 degC at the outlet from the inlet vessel neck and 300 degC from the outlet neck. A total of 100 tests were performed with different outlet opening dimensions and different models of vessel internals fixed in various ways. The starting values mostly corresponded to those for WWER-440 reactors. The results of the experiments indicate a marked thermodynamic off-equilibrium of the process in the starting stage of the outflow and existence of a bonding between the internal structure and the coolant. A semiempirical relation was derived for determining the minimal pressure during the passage of the underpressure expansion wave front. Analysis of the mutual effects of the coolant hydrodynamics and the dynamics of the internals confirmed a dispersion nature of propagation of the exciting impulse in the model. (Z.M.). 4 figs., 3 refs

  20. Modelling of the transfer of CS-137 from air to crops, milk, beef and human body following the Chernobyl accident, in a location in Central Bohemia. Test of the model PRYMA T1

    This work was made in the frame of the research programme on validation of models for the transfer of radionuclides in the terrestrial, urban and aquatic environments. The acronym of this programme is VAMP (Validation of Model Predictions) and is coordinated by the International Atomic Energy Agency (IAEA) and the Commission of the European Communities (CEC). The scenario was named CB and was presented by the Multiple Pathway Working group. The scenario description was at the beginning a blind test, that is without knowing the location or the measured concentrations and doses. The input information included data of contamination in Cs-137 from the Chernobyl accident in Central Europe, in air and soils and more description of the scenario (data about crops, cattele, demography, human diet, etc.). The aim of the exercise was the contrast between model results and between observed data and model predictions. In this work the results obtained by the CIEMAT-IMA group of modelers are shown and discussed

  1. Expert Fuzzy Model for Avalanche Prediction

    Mohan Vizhakat

    2003-10-01

    Full Text Available It is imperative that the time required for the analysis and prediction of an extremely volatile event like avalanche needs to be reduced to the minimum. This is particularly critical because of the extremely fast and highly uncertain nature of the event itself. Another peculiar nature of such predictions is that these have to be based almost entirely on the long and intermediate-term data/infomation available, since there would hardly be any short-term warnings (unlike as in the case of a storm that could point towards an imminent prediction. Both the above-mentioned factors favour adoption of such techniques of automated analysis, which are fast, accurate, and employable even under uncertainvoids of information. Apart from empirical and statistical methods, one of the highly promising techniques for developing a practical model for prediction of avalanche is that based on rule-based expert systems. However, development of a realistic rule-based expert system based on conventional logic would imply that one has to firstly define the natural phenomenon being modelled at an extremely high resolution and accuracy. The process of defining a highly uncertain phenomenon like the avalanche at such high resolution, and thereafter, framing extensive rules for all the possibilities is likely to make the system extremely complex, and therefore, unmanageable in many ways. This study attempts tosimplify this problem by proposing a simpler and better technique using an algorithm based on fuzzy logic. This algorithm has the potential to handle even highly complex phenomenon, like that of an avalanche in a fundamentally simple manner. Such potential makes it capable of handling the higher levels of details and still contains the complexity within the manageable limits. Additional details would also make the system more accurate and realistic.

  2. Predictions in multifield models of inflation

    This paper presents a method for obtaining an analytic expression for the density function of observables in multifield models of inflation with sum-separable potentials. The most striking result is that the density function in general possesses a sharp peak and the location of this peak is only mildly sensitive to the distribution of initial conditions. A simple argument is given for why this result holds for a more general class of models than just those with sum-separable potentials and why for such models, it is possible to obtain robust predictions for observable quantities. As an example, the joint density function of the spectral index and running in double quadratic inflation is computed. For scales leaving the horizon 55 e-folds before the end of inflation, the density function peaks at ns = 0.967 and α = 0.0006 for the spectral index and running respectively

  3. An analytical model for climatic predictions

    A climatic model based upon analytical expressions is presented. This model is capable of making long-range predictions of heat energy variations on regional or global scales. These variations can then be transformed into corresponding variations of some other key climatic parameters since weather and climatic changes are basically driven by differential heating and cooling around the earth. On the basis of the mathematical expressions upon which the model is based, it is shown that the global heat energy structure (and hence the associated climatic system) are characterized by zonally as well as latitudinally propagating fluctuations at frequencies downward of 0.5 day-1. We have calculated the propagation speeds for those particular frequencies that are well documented in the literature. The calculated speeds are in excellent agreement with the measured speeds. (author). 13 refs

  4. Model Predictive Control for Smart Energy Systems

    Halvgaard, Rasmus

    supply electricity reliably to both residential and industrial consumers around the clock. More and more fluctuating renewable energy sources, like wind and solar, are integrated in the power system. Consequently, uncertainty in production starts to affect an otherwise controllable power production...... actors. Chapter 2 provides linear dynamical models of Smart Grid units: Electric Vehicles, buildings with heat pumps, refrigeration systems, solar collectors, heat storage tanks, power plants, and wind farms. The models can be realized as discrete time state space models that fit into a predictive......In this thesis, we consider control strategies for flexible distributed energy resources in the future intelligent energy system – the Smart Grid. The energy system is a large-scale complex network with many actors and objectives in different hierarchical layers. Specifically the power system must...

  5. Modeling the atmospheric dispersion of radioactive effluents in a nuclear accident situation

    In case of a nuclear accident, which could lead to release of radioactive contaminants, fastest countermeasures are needed related to sheltering, iodine distribution, evacuation and interdiction of food and water consumption. All these decisions should be based either on estimation of inhaled dose and the dose due to external exposure for public, or on the estimation of radioactive concentration in food (which will depend on the radioactive concentration in air and ground deposition). The dispersion model used, was a Gaussian 'puff' model. The vertical dispersion was considered not dependent on the release high. The used meteorological data are specific for the SCN - Pitesti site, collected every hour for one year. The meteorological data file contains: the wind speed (in m/s), wind direction (degrees clockwise from north), atmospheric stability category, precipitation rate (in mm/h) and the high of the mixing layer (in m). A hypothetical major nuclear accident at TRIGA - SSR of INR - Pitesti, due to a serious damage of the reactor core leading, to a large release of radioactive contaminants was examined. The release was considered as a single phase with of one hour duration. The release factors for the considered isotopic mixture are 100% noble gases (of the reactor core inventory), 40% iodine (of the reactor core inventory) and 40% particulate, i.e., 40% of the fission products of core fission products inventory, released as particles. The accuracy of the model could be increased by implementation of the code on a real-time system, where the acquisition of the parameters done is on-line, namely, the data are introduced as soon as the modification of meteorological and dosimetric conditions are produced. In this case, the parameters used in formulas can be adjusted according with the field situation. Unfortunately the real-time systems need more powerful resources: monitoring stations which can measure and send on-line the data and which can cover a large area

  6. Efficient optimization for Model Predictive Control in reservoir models

    Borgesen, Jørgen Frenken

    2009-01-01

    The purpose of this thesis was to study the use of adjoint methods for gradient calculations in Model Predictive Control (MPC) applications. The goal was to find and test efficient optimization methods to use in MPC on oil reservoir models. Handling output constraints in the optimization problem has been studied closer since they deteriorate the efficiency of the MPC applications greatly. Adjoint- and finite difference approaches for gradient calculations was tested on reservoir models to de...

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

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

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

    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

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

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

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

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

  11. Permafrost, climate, and change: predictive modelling approach.

    Anisimov, O.

    2003-04-01

    Predicted by GCMs enhanced warming of the Arctic will lead to discernible impacts on permafrost and northern environment. Mathematical models of different complexity forced by scenarios of climate change may be used to predict such changes. Permafrost models that are currently in use may be divided into four groups: index-based models (e.g. frost index model, N-factor model); models of intermediate complexity based on equilibrium simplified solution of the Stephan problem ("Koudriavtcev's" model and its modifications), and full-scale comprehensive dynamical models. New approach of stochastic modelling came into existence recently and has good prospects for the future. Important task is to compare the ability of the models that are different in complexity, concept, and input data requirements to capture the major impacts of changing climate on permafrost. A progressive increase in the depth of seasonal thawing (often referred to as the active-layer thickness, ALT) could be a relatively short-term reaction to climatic warming. At regional and local scales, it may produce substantial effects on vegetation, soil hydrology and runoff, as the water storage capacity of near-surface permafrost will be changed. Growing public concerns are associated with the impacts that warming of permafrost may have on engineered infrastructure built upon it. At the global scale, increase of ALT could facilitate further climatic change if more greenhouse gases are released when the upper layer of the permafrost thaws. Since dynamic permafrost models require complete set of forcing data that is not readily available on the circumpolar scale, they could be used most effectively in regional studies, while models of intermediate complexity are currently best tools for the circumpolar assessments. Set of five transient scenarios of climate change for the period 1980 - 2100 has been constructed using outputs from GFDL, NCAR, CCC, HadCM, and ECHAM-4 models. These GCMs were selected in the course

  12. Model predictive control of smart microgrids

    Hu, Jiefeng; Zhu, Jianguo; Guerrero, Josep M.

    2014-01-01

    required to realise high-performance of distributed generations and will realise innovative control techniques utilising model predictive control (MPC) to assist in coordinating the plethora of generation and load combinations, thus enable the effective exploitation of the clean renewable energy sources......The exploitation of renewable energy and the development of intelligent electricity network have become the main concerns worldwide. This paper aims to integrate renewable energy sources, local loads, and energy storage devices into smart microgrids. It proposes a new microgrid configuration...

  13. Explicit model predictive control accuracy analysis

    Knyazev, Andrew; Zhu, Peizhen; Di Cairano, Stefano

    2015-01-01

    Model Predictive Control (MPC) can efficiently control constrained systems in real-time applications. MPC feedback law for a linear system with linear inequality constraints can be explicitly computed off-line, which results in an off-line partition of the state space into non-overlapped convex regions, with affine control laws associated to each region of the partition. An actual implementation of this explicit MPC in low cost micro-controllers requires the data to be "quantized", i.e. repre...

  14. Model Predictive Control of Wind Turbines

    Henriksen, Lars Christian

    Wind turbines play a major role in the transformation from a fossil fuel based energy production to a more sustainable production of energy. Total-cost-of-ownership is an important parameter when investors decide in which energy technology they should place their capital. Modern wind turbines are...... been suggested as an alternative to ground-fixed wind turbines as they can be placed at water depths usually thought outside the realm of wind turbine placement. The special challenges posed by controlling a floating wind turbine have been addressed in this thesis. Model predictive control (MPC) has...

  15. Distributed Model Predictive Control via Dual Decomposition

    Biegel, Benjamin; Stoustrup, Jakob; Andersen, Palle

    2014-01-01

    This chapter presents dual decomposition as a means to coordinate a number of subsystems coupled by state and input constraints. Each subsystem is equipped with a local model predictive controller while a centralized entity manages the subsystems via prices associated with the coupling constraints....... This allows coordination of all the subsystems without the need of sharing local dynamics, objectives and constraints. To illustrate this, an example is included where dual decomposition is used to resolve power grid congestion in a distributed manner among a number of players coupled by distribution...

  16. Measurements and modelling of 137Cs distribution on ground due to the Chernobyl accident: a 27-y follow-up study in Northern Greece

    Following the Chernobyl accident, an area of ∼1000 m2 in the University farm of the Aristotle University of Thessaloniki was considered as a test ground for radioecological measurements. The radiocesium deposition in this area, due to the Chernobyl accident, was 20 kBq m-2. The profile of 137Cs in the soil of this area was measured systematically from 1987 to 2012. The form of the profile has changed over the years. During the 1987-2000 period the 137Cs distribution was reproducible by a sum of two exponentials. However, at least since 2005 the 137Cs distribution can be successfully fitted by a single exponential function. The long-time (∼27 y) evolution study of the 137Cs distribution in soil permit one to extract with the use of a simple compartment model, the mean vertical migration velocity of 137Cs. Vertical migration of 137Cs in soil is a very slow process. The mean vertical migration velocity is estimated to be 0.14 cm y-1.The relative good comparison between the time dependence of the 137Cs distribution in soil and the model predictions indicate that the simple model used is realistic. (authors)

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

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

  18. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Massih, Ali R. [Quantum Technologies AB, Uppsala Science Park (Sweden)

    2007-02-15

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made.

  19. Applying the DSNP modular modeling system to transient and accident simulations of lead cooled reactors

    The modeling and simulation of the Encapsulated Nuclear Heat Source (ENHS) is presented in this study. The purpose of the simulations was to evaluate the safety characteristics of the proposed modular liquid metal cooled reactor. The DSNP simulation package was modified to accept LBE (Lead Bismuth Eutectic) and lead as a reactor coolant and as heat transfer medium in the primary and secondary loops. Appropriate equations of state, heat transfer and flow correlations were also introduced to permit a full range of simulations of the ENHS and other lead and LBE cooled systems. Models of different levels of complexity were developed to study various events and their consequences. Due to the very large heat capacity of the ENHS reactor, unusually long simulation times ranging from hours up to days were needed to follow some of the transients. This in turn required modifications to various elements of the DSNP simulation system to permit these long execution times. It is concluded that the ENHS has an inherently safe response to all initiating events, and that the DSNP system is capable to simulate most of the accidents of interest to the safety evaluation of the plant. (author)

  20. Review of experimental data for modelling LWR fuel cladding behaviour under loss of coolant accident conditions

    Extensive range of experiments has been conducted in the past to quantitatively identify and understand the behaviour of fuel rod under loss-of-coolant accident (LOCA) conditions in light water reactors (LWRs). The obtained experimental data provide the basis for the current emergency core cooling system acceptance criteria under LOCA conditions for LWRs. The results of recent experiments indicate that the cladding alloy composition and high burnup effects influence LOCA acceptance criteria margins. In this report, we review some past important and recent experimental results. We first discuss the background to acceptance criteria for LOCA, namely, clad embrittlement phenomenology, clad embrittlement criteria (limitations on maximum clad oxidation and peak clad temperature) and the experimental bases for the criteria. Two broad kinds of test have been carried out under LOCA conditions: (i) Separate effect tests to study clad oxidation, clad deformation and rupture, and zirconium alloy allotropic phase transition during LOCA. (ii) Integral LOCA tests, in which the entire LOCA sequence is simulated on a single rod or a multi-rod array in a fuel bundle, in laboratory or in a tests and results are discussed and empirical correlations deduced from these tests and quantitative models are conferred. In particular, the impact of niobium in zirconium base clad and hydrogen content of the clad on allotropic phase transformation during LOCA and also the burst stress are discussed. We review some recent LOCA integral test results with emphasis on thermal shock tests. Finally, suggestions for modelling and further evaluation of certain experimental results are made