Analysis of Fukushima Daiichi Accident Using HFACS

International Nuclear Information System (INIS)

Mohamed, Saeed Almheiri

2013-01-01

The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO 1 and NISA 2 that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident

Analysis of Fukushima Daiichi Accident Using HFACS

Energy Technology Data Exchange (ETDEWEB)

Mohamed, Saeed Almheiri [Korea Advanced Institue of Science and Technology, Daejeon (Korea, Republic of)

2013-10-15

The shadow of Fukushima Daiichi nuclear power plant (NPP) accident is still too big and will last long. On the other hand, it could still teach us lots of lessons to better design and operate nuclear power plants. In this paper, we will be focusing on the Fukushima Daiichi accident, especially on human organizational factors. We will analyze the accident using Human Factors Analysis and Classification System (HFACS) in order to better understand the organizational climate of TEPCO{sup 1} and NISA{sup 2} that led to Fukushima Daiichi Accident. HFACS was developed for the U. S. aviation industry and has been used at many industries like the rail and mining industries. We found that the HFACS to be greatly beneficial in investigating the latent and organizational causes for the accident. The application results show that the causes of Fukushima Daiichi accident were spread out from sharp end (i.e. Unsafe Act) to blunt end (i. e. Organizational Influences). This means that the corresponding countermeasures should cover from front line staff to management. Thus, we managed to develop a better understanding on how to prevent similar errors or violations. The incident and near-miss have a lot of helpful information because it may show the actual and latent deficiencies of complex systems. We applied the HFACS into Fukushima Daiichi accident to better locate the causes related to both sharp and blunt ends of operation of NPP. In order to derive useful lessons from the accident analysis, the analyst should try to find the similarities not differences from the incident. It is imperative that whatever accident/incident analysis systems we use, we should fully utilize the disastrous Fukushima accident.

Qualitative analysis of the man-organization system in accident conditions for nuclear installations

International Nuclear Information System (INIS)

Farcasiu, Mita; Prisecaru, Ilie

2010-01-01

In this paper a model of the human performance investigation of accident conditions in the operation of the nuclear installation is developed. A framework for analyses of the human action in the man-organization system context is achieved. The goal of this model is to identify the possible roots causing human errors which could occur during the evolution of the accident by the qualitative analysis of the interfaces in man-organization system. These interfaces represent the main elements which characterize the implication of the organization in human performance. The results of this paper are the interfaces of the man-organization and their circumstances in which human performance could fail. Also, another result is a pre-designed framework which could help in the investigation of an accident. (authors)

Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

Energy Technology Data Exchange (ETDEWEB)

WILLIAMS, J.C.

2000-09-15

Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR).

Analysis of Waste Leak and Toxic Chemical Release Accidents from Waste Feed Delivery (WFD) Diluent System

International Nuclear Information System (INIS)

WILLIAMS, J.C.

2000-01-01

Radiological and toxicological consequences are calculated for 4 postulated accidents involving the Waste Feed Delivery (WFD) diluent addition systems. Consequences for the onsite and offsite receptor are calculated. This analysis contains technical information used to determine the accident consequences for the River Protection Project (RPP) Final Safety Analysis Report (FSAR)

An Accident Precursor Analysis Process Tailored for NASA Space Systems

Science.gov (United States)

Groen, Frank; Stamatelatos, Michael; Dezfuli, Homayoon; Maggio, Gaspare

2010-01-01

Accident Precursor Analysis (APA) serves as the bridge between existing risk modeling activities, which are often based on historical or generic failure statistics, and system anomalies, which provide crucial information about the failure mechanisms that are actually operative in the system and which may differ in frequency or type from those in the various models. These discrepancies between the models (perceived risk) and the system (actual risk) provide the leading indication of an underappreciated risk. This paper presents an APA process developed specifically for NASA Earth-to-Orbit space systems. The purpose of the process is to identify and characterize potential sources of system risk as evidenced by anomalous events which, although not necessarily presenting an immediate safety impact, may indicate that an unknown or insufficiently understood risk-significant condition exists in the system. Such anomalous events are considered accident precursors because they signal the potential for severe consequences that may occur in the future, due to causes that are discernible from their occurrence today. Their early identification allows them to be integrated into the overall system risk model used to intbrm decisions relating to safety.

Causal Analysis to a Subway Accident: A Comparison of STAMP and RAIB

Directory of Open Access Journals (Sweden)

Zhou Yao

2018-01-01

Full Text Available Accident investigation and analysis after the accident, vital to prevent the occurrence of similar accident and improve the safety of the system. Different methods led to a different understanding of the accident. In this paper, a subway accident was analysed with a systemic accident analysis model – STAMP (System-Theoretic Accident Modelling and Processes. The hierarchical safety control structure was obtained, and the system-level safety constraints were obtained, controllers of the physical layer were analysed one by one, and put forward the relevant safety requirements and constraints, the dynamic analysis of the structure of the safety control is carried out, and the targeted recommendations are pointed out. In comparison with the analysis results obtained by the Rail Accident Investigation Branch (RAIB. Some useful findings have been concluded. STAMP treats safety as a control problem and reduces or eliminates causes of the accident from the controlling perspective. Whereas RAIB obtains causes of the accident by analysing the sequence of events related to the accident and reasons of these events, then chooses one(or moreevent(s as the immediate cause and some of the key events as causal factors. RAIB analysis is based on the sequential event models, but STAMP analysis provides us with a holistic, dynamic way to control system to maintain safety.

Development of the severe accident risk information database management system SARD

International Nuclear Information System (INIS)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

Development of the severe accident risk information database management system SARD

Energy Technology Data Exchange (ETDEWEB)

Ahn, Kwang Il; Kim, Dong Ha

2003-01-01

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

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.

1985-01-01

SAS4A is a new code system which has been designed for analyzing the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modeling the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel motion experiment analyses are also presented

Modification of MELCOR for severe accident analysis of candidate accident tolerant cladding materials

Energy Technology Data Exchange (ETDEWEB)

Merrill, Brad J., E-mail: brad.merrill@inl.gov; Bragg-Sitton, Shannon M., E-mail: shannon.bragg-sitton@inl.gov; Humrickhouse, Paul W., E-mail: paul.humrickhouse@inl.gov

2017-04-15

Highlights: • Accident tolerant fuels (ATF) systems are currently under development for LWRs. • Many performance analysis tools are specifically developed for UO{sub 2}–Zr alloy fuel. • Modifications were made to the MELCOR code for candidate ATF cladding. • Preliminary analysis results for SiC and FeCrAl cladding concepts are presented. - Abstract: A number of materials are currently under development as candidate accident tolerant fuel and cladding for application in the current fleet of commercial light water reactors (LWRs). The safe, reliable and economic operation of the nation’s nuclear power reactor fleet has always been a top priority for the nuclear industry. Continual improvement of technology, including advanced materials and nuclear fuels, remains central to the industry’s success. Enhancing the accident tolerance of light water reactors became a topic of serious discussion following the 2011 Great East Japan Earthquake, resulting tsunami, and subsequent damage to the Fukushima Daiichi nuclear power plant complex. The overall goal for the development of accident tolerant fuel (ATF) systems for LWRs is to identify alternative fuel system technologies to further enhance the safety, competitiveness, and economics of commercial nuclear power. Designed for use in the current fleet of commercial LWRs, or in reactor concepts with design certifications (GEN-III+), to achieve their goal enhanced ATF must endure loss of active cooling in the reactor core for a considerably longer period of time than the current fuel system, while maintaining or improving performance during normal operation. Many available nuclear fuel performance analysis tools are specifically developed for the current UO{sub 2}–Zirconium alloy fuel system. The MELCOR severe-accident analysis code, under development at the Sandia National Laboratory in New Mexico (SNL-NM) for the US Nuclear Regulatory Commission (NRC), is one of these tools. This paper describes modifications

Accident Analysis and Barrier Function (AEB) Method. Manual for Incident Analysis

International Nuclear Information System (INIS)

Svenson, Ola

2000-02-01

The Accident Analysis and Barrier Function (AEB) Method models an accident or incident as a series of interactions between human and technical systems. In the sequence of human and technical errors leading to an accident there is, in principle, a possibility to arrest the development between each two successive errors. This can be done by a barrier function which, for example, can stop an operator from making an error. A barrier function can be performed by one or several barrier function systems. To illustrate, a mechanical system, a computer system or another operator can all perform a given barrier function to stop an operator from making an error. The barrier function analysis consists of analysis of suggested improvements, the effectiveness of the improvements, the costs of implementation, probability of implementation, the cost of maintaining the barrier function, the probability that maintenance will be kept up to standards and the generalizability of the suggested improvement. The AEB method is similar to the US method called HPES, but differs from that method in different ways. To exemplify, the AEB method has more emphasis on technical errors than HPES. In contrast to HPES that describes a series of events, the AEB method models only errors. This gives a more focused analysis making it well suited for checking other HPES-type accident analyses. However, the AEB method is a generic and stand-alone method that has been applied in other fields than nuclear power, such as, in traffic accident analyses

Accident Analysis and Barrier Function (AEB) Method. Manual for Incident Analysis

Energy Technology Data Exchange (ETDEWEB)

Svenson, Ola [Stockholm Univ. (Sweden). Dept. of Psychology

2000-02-01

The Accident Analysis and Barrier Function (AEB) Method models an accident or incident as a series of interactions between human and technical systems. In the sequence of human and technical errors leading to an accident there is, in principle, a possibility to arrest the development between each two successive errors. This can be done by a barrier function which, for example, can stop an operator from making an error. A barrier function can be performed by one or several barrier function systems. To illustrate, a mechanical system, a computer system or another operator can all perform a given barrier function to stop an operator from making an error. The barrier function analysis consists of analysis of suggested improvements, the effectiveness of the improvements, the costs of implementation, probability of implementation, the cost of maintaining the barrier function, the probability that maintenance will be kept up to standards and the generalizability of the suggested improvement. The AEB method is similar to the US method called HPES, but differs from that method in different ways. To exemplify, the AEB method has more emphasis on technical errors than HPES. In contrast to HPES that describes a series of events, the AEB method models only errors. This gives a more focused analysis making it well suited for checking other HPES-type accident analyses. However, the AEB method is a generic and stand-alone method that has been applied in other fields than nuclear power, such as, in traffic accident analyses.

System response of a DOE Defense Program package in a transportation accident environment

International Nuclear Information System (INIS)

Chen, T.F.; Hovingh, J.; Kimura, C.Y.

1992-01-01

The system response in a transportation accident environment is an element to be considered in an overall Transportation System Risk Assessment (TSRA) framework. The system response analysis uses the accident conditions and the subsequent accident progression analysis to develop the accident source term, which in turn, is used in the consequence analysis. This paper proposes a methodology for the preparation of the system response aspect of the TSRA

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

Park, Chul; Kim, Heon Il; Lee, Bo Yook; Lee, Sang Yong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

For the safety assessment of HANARO, accident analyses for the anticipated operational transients, accident scenarios and limiting accident scenarios were conducted. To do this, the commercial nuclear reactor system code. RELAP5/MOD2 was modified to RELAP5/KMRR; the thermal hydraulic correlations and the heat exchanger model was changed to incorporate HANARO characteristics. This report summarizes the RELAP/KMRR calculation results and the subchannel analyses results based on the RELAP/KMRR results. During the calculation, major concern was placed on the integrity of the fuel. For all the scenarios, the important accident analysis parameters, i.e., fuel centerline temperatures and the minimum critical heat flux ratio(MCHFR), satisfied safe design limits. It was verified, therefore, that the HANARO was safely designed. 21 tabs., 89 figs., 39 refs. (Author) .new.

Accident analysis of HANARO fuel test loop

Energy Technology Data Exchange (ETDEWEB)

Kim, J. Y.; Chi, D. Y

1998-03-01

Steady state fuel test loop will be equipped in HANARO to obtain the development and betterment of advanced fuel and materials through the irradiation tests. The HANARO fuel test loop was designed to match the CANDU and PWR fuel operating conditions. The accident analysis was performed by RELAP5/MOD3 code based on FTL system designs and determined the detail engineering specification of in-pile test section and out-pile systems. The accident analysis results of FTL system could be used for the fuel and materials designer to plan the irradiation testing programs. (author). 23 refs., 20 tabs., 178 figs.

Post-accident cooling capacity analysis of the AP1000 passive spent fuel pool cooling system

International Nuclear Information System (INIS)

Su Xia

2013-01-01

The passive design is used in AP1000 spent fuel pool cooling system. The decay heat of the spent fuel is removed by heating-boiling method, and makeup water is provided passively and continuously to ensure the safety of the spent fuel. Based on the analysis of the post-accident cooling capacity of the spent fuel cooling system, it is found that post-accident first 72-hour cooling under normal refueling condition and emergency full-core offload condition can be maintained by passive makeup from safety water source; 56 hours have to be waited under full core refueling condition to ensure the safety of the core and the spent fuel pool. Long-term cooling could be conducted through reserved safety interface. Makeup measure is available after accident and limited operation is needed. Makeup under control could maintain the spent fuel at sub-critical condition. Compared with traditional spent fuel pool cooling system design, the AP1000 design respond more effectively to LOCA accidents. (authors)

analysis of reactivity accidents in MTR for various protection system parameters and core condition

International Nuclear Information System (INIS)

Mohamed, F.M.

2011-01-01

Egypt Second Research Reactor (ETRR-2) core was modified to irradiate LEU (Low Enriched Uranium) plates in two irradiation boxes for fission 99 Mo production. The old core comprising 29 fuel elements and one Co Irradiation Device (CID) and the new core comprising 27 fuel elements, CID, and two 99 Mo production boxes. The in core irradiation has the advantage of no special cooling or irradiation loop is required. The purpose of the present work is the analysis of reactivity accidents (RIA) for ETRR-2 cores. The analysis was done to evaluate the accidents from different point of view:1- Analysis of the new core for various Reactor Protection System (RPS) parameters 2- Comparison between the two cores. 3- Analysis of the 99 Mo production boxes.PARET computer code was employed to compute various parameters. Initiating events in RIA involve various modes of reactivity insertion, namely, prompt critical condition (p=1$), accidental ejection of partial and complete CID uncontrolled withdrawal of a control rod accident, and sudden cooling of the reactor core. The time histories of reactor power, energy released, and the maximum fuel, clad and coolant temperatures of fuel elements and LEU plates were calculated for each of these accidents. The results show that the maximum clad temperatures remain well below the clad melting of both fuel and uranium plates during these accidents. It is concluded that for the new core, the RIA with scram will not result in fuel or uranium plate failure.

Probabilistic Accident Progression Analysis with application to a LMFBR design

International Nuclear Information System (INIS)

Jamali, K.M.

1982-01-01

A method for probabilistic analysis of accident sequences in nuclear power plant systems referred to as ''Probabilistic Accident Progression Analysis'' (PAPA) is described. Distinctive features of PAPA include: (1) definition and analysis of initiator-dependent accident sequences on the component level; (2) a new fault-tree simplification technique; (3) a new technique for assessment of the effect of uncertainties in the failure probabilities in the probabilistic ranking of accident sequences; (4) techniques for quantification of dependent failures of similar components, including an iterative technique for high-population components. The methodology is applied to the Shutdown Heat Removal System (SHRS) of the Clinch River Breeder Reactor Plant during its short-term (0 -2 . Major contributors to this probability are the initiators loss of main feedwater system, loss of offsite power, and normal shutdown

Severe accident analysis methodology in support of accident management

International Nuclear Information System (INIS)

Boesmans, B.; Auglaire, M.; Snoeck, J.

1997-01-01

The author addresses the implementation at BELGATOM of a generic severe accident analysis methodology, which is intended to support strategic decisions and to provide quantitative information in support of severe accident management. The analysis methodology is based on a combination of severe accident code calculations, generic phenomenological information (experimental evidence from various test facilities regarding issues beyond present code capabilities) and detailed plant-specific technical information

Accident Sequence Evaluation Program: Human reliability analysis procedure

International Nuclear Information System (INIS)

Swain, A.D.

1987-02-01

This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs

Chernobyl - system accident or human error?

International Nuclear Information System (INIS)

Stang, E.

1996-01-01

Did human error cause the Chernobyl disaster? The standard point of view is that operator error was the root cause of the disaster. This was also the view of the Soviet Accident Commission. The paper analyses the operator errors at Chernobyl in a system context. The reactor operators committed errors that depended upon a lot of other failures that made up a complex accident scenario. The analysis is based on Charles Perrow's analysis of technological disasters. Failure possibility is an inherent property of high-risk industrial installations. The Chernobyl accident consisted of a chain of events that were both extremely improbable and difficult to predict. It is not reasonable to put the blame for the disaster on the operators. (author)

Analysis of three loss-of-flow accidents in the first wall cooling system of NET/ITER

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1993-05-01

This report presents the thermal-hydraulic analysis of three Loss-of-Flow Accidents (LOFAs) in the first wall cooling system of the Next European Torus (NET) design or the International Thermonuclear Experimental Reactor (ITER) design. The LOFAs considered result from a loss of the forced coolant flow caused by a loss of electrical power for the recirculation pump in the primary circuit. The analyses have been performed using the thermal-hydraulic system analysis code RELAP5/MOD3. In the analyses, special attention has been paid to the transient thermal-hydraulic behaviour of the cooling system and the temperature development in the first wall. In the LOFA case without plasma shutdown, melting starts in the first wall about 150 s after accident initiation. In the LOFA case with delayed plasma shutdown, melting starts in the first wall when the plasma shutdown is initiated later than about 110 s after accident initiation. Melting does not occur in the first wall during a LOFA with prompt plasma scram. (orig.)

Accident Sequence Precursor Analysis for SGTR by Using Dynamic PSA Approach

International Nuclear Information System (INIS)

Lee, Han Sul; Heo, Gyun Young; Kim, Tae Wan

2016-01-01

In order to address this issue, this study suggests the sequence tree model to analyze accident sequence systematically. Using the sequence tree model, all possible scenarios which need a specific safety action to prevent the core damage can be identified and success conditions of safety action under complicated situation such as combined accident will be also identified. Sequence tree is branch model to divide plant condition considering the plant dynamics. Since sequence tree model can reflect the plant dynamics, arising from interaction of different accident timing and plant condition and from the interaction between the operator action, mitigation system, and the indicators for operation, sequence tree model can be used to develop the dynamic event tree model easily. Target safety action for this study is a feed-and-bleed (F and B) operation. A F and B operation directly cools down the reactor cooling system (RCS) using the primary cooling system when residual heat removal by the secondary cooling system is not available. In this study, a TLOFW accident and a TLOFW accident with LOCA were the target accidents. Based on the conventional PSA model and indicators, the sequence tree model for a TLOFW accident was developed. Based on the results of a sampling analysis and data from the conventional PSA model, the CDF caused by Sequence no. 26 can be realistically estimated. For a TLOFW accident with LOCA, second accident timings were categorized according to plant condition. Indicators were selected as branch point using the flow chart and tables, and a corresponding sequence tree model was developed. If sampling analysis is performed, practical accident sequences can be identified based on the sequence analysis. If a realistic distribution for the variables can be obtained for sampling analysis, much more realistic accident sequences can be described. Moreover, if the initiating event frequency under a combined accident can be quantified, the sequence tree model

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

Energy Technology Data Exchange (ETDEWEB)

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

1992-01-15

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

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

International Nuclear Information System (INIS)

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

1992-01-01

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

URBAN TRAFFIC ACCIDENT ANALYSIS BY USING GEOGRAPHIC INFORMATION SYSTEM

Directory of Open Access Journals (Sweden)

Meltem SAPLIOĞLU

2006-03-01

Full Text Available In recent years, traffic accidents that cause more social and economic losses than that of natural disasters,have become a national problem in Turkey. To solve this problem and to reduce the casualties, road safety programs are tried to be developed. It is necessary to develop the most effective measures with low investment cost due to limited budgets allocated to such road safety programs. The most important program is to determine dangerous locations of traffic accidents and to improve these sections from the road safety view point. New Technologies are driving a cycle of continuous improvement that causes rapid changes in the traffic engineering and any engineering services within it. It is obvious that this developed services will be the potential for forward-thinking engineering studies to take a more influence role. In this study, Geographic Information System (GIS was used to identify the hazardous locations of traffic accidents in Isparta. Isparta city map was digitized by using Arcinfo 7.21. Traffic accident reports occurred between 1998-2002 were obtained from Directory of Isparta Traffic Region and had been used to form the database. Topology was set up by using Crash Diagrams and Geographic Position Reference Systems. Tables are formed according to the obtained results and interpreted.

Review of severe accidents and the results of accident consequence assessment in different energy systems (Contract research)

International Nuclear Information System (INIS)

Matsuki, Yoshio; Muramatsu, Ken

2008-05-01

The cases of severe accidents and the consequence assessments in different energy systems, Coal, Oil, Gas, Hydro and Nuclear, were collected, and then they were further analyzed. In this report, the information on the accidents in various energy systems were collected from the sources of the Paul Scherrer Institute (hereinafter, 'PSI') and the International Atomic Energy Agency (hereinafter, 'IAEA'). The information on the severe accidents of nuclear power plants were collected from the report of the US Presidential Commission on Catastrophic Nuclear Accidents and several relevant reports issued in the countries of the European Union, together with the reports of the PSI and the IAEA. To analyze the collected information, several parameters, which are numbers of fatalities, injuries, evacuees and the costs of the damages, were chosen to characterize those accidents in different energy systems. And then, upon the comparison of these characteristics of different accidents, the impacts of the accidents in nuclear and other energy systems were compared. Upon the results of the analysis, it is pointed out that the cost caused by the Chernobyl Accident, the severe accident in nuclear energy, tends to be higher than in the other energy systems. On the other hand, from the aspects of fatalities and injuries, it is not confirmed that the damages of the Chernobyl Accident are larger than in the other energy systems. However, it is also recognized, as the specific characteristics of the severe nuclear accident, that the impacts of the accident spread in a wider area, and stay for a longer period, in comparison with the ones in the other energy systems. (author)

Accident Sequence Evaluation Program: Human reliability analysis procedure

Energy Technology Data Exchange (ETDEWEB)

Swain, A.D.

1987-02-01

This document presents a shortened version of the procedure, models, and data for human reliability analysis (HRA) which are presented in the Handbook of Human Reliability Analysis With emphasis on Nuclear Power Plant Applications (NUREG/CR-1278, August 1983). This shortened version was prepared and tried out as part of the Accident Sequence Evaluation Program (ASEP) funded by the US Nuclear Regulatory Commission and managed by Sandia National Laboratories. The intent of this new HRA procedure, called the ''ASEP HRA Procedure,'' is to enable systems analysts, with minimal support from experts in human reliability analysis, to make estimates of human error probabilities and other human performance characteristics which are sufficiently accurate for many probabilistic risk assessments. The ASEP HRA Procedure consists of a Pre-Accident Screening HRA, a Pre-Accident Nominal HRA, a Post-Accident Screening HRA, and a Post-Accident Nominal HRA. The procedure in this document includes changes made after tryout and evaluation of the procedure in four nuclear power plants by four different systems analysts and related personnel, including human reliability specialists. The changes consist of some additional explanatory material (including examples), and more detailed definitions of some of the terms. 42 refs.

A systemic analysis of South Korea Sewol ferry accident - Striking a balance between learning and accountability.

Science.gov (United States)

Kee, Dohyung; Jun, Gyuchan Thomas; Waterson, Patrick; Haslam, Roger

2017-03-01

The South Korea Sewol ferry accident in April 2014 claimed the lives of over 300 passengers and led to criminal charges of 399 personnel concerned including imprisonment of 154 of them as of Oct 2014. Blame and punishment culture can be prevalent in a more hierarchical society like South Korea as shown in the aftermath of this disaster. This study aims to analyse the South Korea ferry accident using Rasmussen's risk management framework and the associated AcciMap technique and to propose recommendations drawn from an AcciMap-based focus group with systems safety experts. The data for the accident analysis were collected mainly from an interim investigation report by the Board of Audit and Inspection of Korea and major South Korean and foreign newspapers. The analysis showed that the accident was attributed to many contributing factors arising from front-line operators, management, regulators and government. It also showed how the multiple factors including economic, social and political pressures and individual workload contributed to the accident and how they affected each other. This AcciMap was presented to 27 safety researchers and experts at 'the legacy of Jens Rasmussen' symposium adjunct to ODAM2014. Their recommendations were captured through a focus group. The four main recommendations include forgive (no blame and punishment on individuals), analyse (socio-technical system-based), learn (from why things do not go wrong) and change (bottom-up safety culture and safety system management). The findings offer important insights into how this type of accident should be understood, analysed and the subsequent response. Copyright © 2016 Elsevier Ltd. All rights reserved.

[A monitoring system for work-related accidents in Piracicaba, São Paulo, Brazil].

Science.gov (United States)

Cordeiro, Ricardo; Vilela, Rodolfo Andrade Gouveia; de Medeiros, Maria Angélica Tavares; Gonçalves, Cláudia Giglio de Oliveira; Bragantini, Clarice Aparecida; Varolla, Antenor J; Celso, Stephan

2005-01-01

The authors report on the development of a work accident monitoring system in Piracicaba, São Paulo State, Brazil, with the following characteristics: information feeding the system is obtained in real time directly from accident treatment centers; the system has universal monitoring, covering all work-related accidents in Piracicaba, regardless of the nature of the worker's employment conditions, place of work, or place of residence; health surveillance and promotion of health initiatives are triggered by identification of sentinel events; spatial distribution analysis of work-related accidents is a basic tool in designing accident awareness strategies and accident prevention policies. The system was implemented in November 2003 and by October 2004 had identified 5,320 work-related accidents, or a 3.8% annual proportional incidence of work-related accidents in the municipal area. We illustrate spatial analysis of registered work-related accidents and present a detailed investigation of one example of a serious accident.

Severe Accident Management System On-line Network SAMSON

International Nuclear Information System (INIS)

Silverman, Eugene B.

2004-01-01

SAMSON is a computational tool used by accident managers in the Technical Support Centers (TSC) and Emergency Operations Facilities (EOF) in the event of a nuclear power plant accident. SAMSON examines over 150 status points monitored by nuclear power plant process computers during a severe accident and makes predictions about when core damage, support plate failure, and reactor vessel failure will occur. These predictions are based on the current state of the plant assuming that all safety equipment not already operating will fail. SAMSON uses expert systems, as well as neural networks trained with the back propagation learning algorithms to make predictions. Training on data from an accident analysis code (MAAP - Modular Accident Analysis Program) allows SAMSON to associate different states in the plant with different times to critical failures. The accidents currently recognized by SAMSON include steam generator tube ruptures (SGTRs), with breaks ranging from one tube to eight tubes, and loss of coolant accidents (LOCAs), with breaks ranging from 0.0014 square feet (1.30 cm 2 ) in size to breaks 3.0 square feet in size (2800 cm 2 ). (author)

Current status of accident analysis for Korean HCCR TBS

Energy Technology Data Exchange (ETDEWEB)

Ahn, Mu-Young, E-mail: myahn74@nfri.re.kr [National Fusion Research Institute, Daejeon (Korea, Republic of); Jin, Hyung Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Cho, Seungyon [National Fusion Research Institute, Daejeon (Korea, Republic of); Lee, Dong Won [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin [National Fusion Research Institute, Daejeon (Korea, Republic of)

2014-10-15

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements.

Current status of accident analysis for Korean HCCR TBS

International Nuclear Information System (INIS)

Ahn, Mu-Young; Jin, Hyung Gon; Cho, Seungyon; Lee, Dong Won; Ku, Duck Young; Park, Yi-Hyun; Kim, Chang-Shuk; Lee, Youngmin

2014-01-01

Korea has decided to test Helium Cooled Ceramic Reflector (HCCR) Test Blanket Module (TBM) in ITER and design of the TBM with its ancillary systems, i.e. Test Blanket System (TBS), is under progress. Since the TBM is operated at elevated temperature with high heat load, safety consideration is essential in design procedure. In this paper, preliminary accident analysis results for the current HCCR TBS design on selected scenarios are presented as an important part of safety assessments. To simulate transient thermo-hydraulic behavior, GAMMA-FR code which has been developed in Korea for fusion applications was used. The main cooling and tritium extraction circuit systems, as well as the TBM, were simulated and the main components in the TBS were modeled as the associated heat structures. The important accident scenarios were produced and summarized in the paper considering the HCCR TBS design and ITER conditions, which cover in-vessel Loss Of Coolant Accident (LOCA), in-box LOCA, ex-vessel LOCA, Loss Of Flow Accident (LOFA), Loss Of Heat Sink Accident (LOHSA) and purge pipe rupture case. The accident analysis based on the selected scenarios was performed and it was found that the current design of the HCCR TBS meets the thermo-hydraulic safety requirements

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

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

A Review of Accident Modelling Approaches for Complex Critical Sociotechnical Systems

National Research Council Canada - National Science Library

Qureshi, Zahid H

2008-01-01

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

Incorporation of advanced accident analysis methodology into safety analysis reports

International Nuclear Information System (INIS)

2003-05-01

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

Analysis of severe core damage accident progression for the heavy water reactor

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

In this study, the severe accident progression analysis of generic Canadian deuterium uranium reactor 6 was preliminarily provided using an integrated severe accident analysis code. The selected accident sequences were multiple steam generator tube rupture and large break loss-of-coolant accidents because these led to severe core damage with an assumed unavailability for several critical safety systems. The progressions of severe accident included a set of failed safety systems normally operated at full power, and initiative events led to primary heat transport system inventory blow-down or boil off. The core heat-up and melting, steam generator response,fuel channel and calandria vessel failure were analyzed. The results showed that the progression of a severe core damage accident induced by steam generator tube rupture or large break loss-of-coolant accidents in a CANDU reactor was slow due to heat sinks in the calandria vessel and vault. (authors)

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

Energy Technology Data Exchange (ETDEWEB)

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

2004-01-01

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

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Analysis of severe accidents in pressurized heavy water reactors

International Nuclear Information System (INIS)

2008-06-01

Certain very low probability plant states that are beyond design basis accident conditions and which may arise owing to multiple failures of safety systems leading to significant core degradation may jeopardize the integrity of many or all the barriers to the release of radioactive material. Such event sequences are called severe accidents. It is required in the IAEA Safety Requirements publication on Safety of the Nuclear Power Plants: Design, that consideration be given to severe accident sequences, using a combination of engineering judgement and probabilistic methods, to determine those sequences for which reasonably practicable preventive or mitigatory measures can be identified. Acceptable measures need not involve the application of conservative engineering practices used in setting and evaluating design basis accidents, but rather should be based on realistic or best estimate assumptions, methods and analytical criteria. Recently, the IAEA developed a Safety Report on Approaches and Tools for Severe Accident Analysis. This publication provides a description of factors important to severe accident analysis, an overview of severe accident phenomena and the current status in their modelling, categorization of available computer codes, and differences in approaches for various applications of severe accident analysis. The report covers both the in- and ex-vessel phases of severe accidents. The publication is consistent with the IAEA Safety Report on Accident Analysis for Nuclear Power Plants and can be considered as a complementary report specifically devoted to the analysis of severe accidents. Although the report does not explicitly differentiate among various reactor types, it has been written essentially on the basis of available knowledge and databases developed for light water reactors. Therefore its application is mostly oriented towards PWRs and BWRs and, to a more limited extent, they can be only used as preliminary guidance for other types of reactors

Analysis on the Role of RSG-GAS Pool Cooling System during Partial Loss of Heat Sink Accident

Science.gov (United States)

Susyadi; Endiah, P. H.; Sukmanto, D.; Andi, S. E.; Syaiful, B.; Hendro, T.; Geni, R. S.

2018-02-01

RSG-GAS is a 30 MW reactor that is mostly used for radioisotope production and experimental activities. Recently, it is regularly operated at half of its capacity for efficiency reason. During an accident, especially loss of heat sink, the role of its pool cooling system is very important to dump decay heat. An analysis using single failure approach and partial modeling of RELAP5 performed by S. Dibyo, 2010 shows that there is no significant increase in the coolant temperature if this system is properly functioned. However lessons learned from the Fukushima accident revealed that an accident can happen due to multiple failures. Considering ageing of the reactor, in this research the role of pool cooling system is to be investigated for a partial loss of heat sink accident which is at the same time the protection system fails to scram the reactor when being operated at 15 MW. The purpose is to clarify the transient characteristics and the final state of the coolant temperature. The method used is by simulating the system in RELAP5 code. Calculation results shows the pool cooling systems reduce coolant temperature for about 1 K as compared without activating them. The result alsoreveals that when the reactor is being operated at half of its rated power, it is still in safe condition for a partial loss of heat sink accident without scram.

Reactivity insertion accident analysis

International Nuclear Information System (INIS)

Moreira, J.M.L.; Nakata, H.; Yorihaz, H.

1990-04-01

The correct prediction of postulated accidents is the fundamental requirement for the reactor licensing procedures. Accident sequences and severity of their consequences depend upon the analysis which rely on analytical tools which must be validated against known experimental results. Present work presents a systematic approach to analyse and estimate the reactivity insertion accident sequences. The methodology is based on the CINETHICA code which solves the point-kinetics/thermohydraulic coupled equations with weighted temperature feedback. Comparison against SPERT experimental results shows good agreement for the step insertion accidents. (author) [pt

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

International Nuclear Information System (INIS)

Naitoh, Masanori; Ujita, Hiroshi; Nagumo, Hiroichi

1997-01-01

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

Accident sequence analysis of human-computer interface design

International Nuclear Information System (INIS)

Fan, C.-F.; Chen, W.-H.

2000-01-01

It is important to predict potential accident sequences of human-computer interaction in a safety-critical computing system so that vulnerable points can be disclosed and removed. We address this issue by proposing a Multi-Context human-computer interaction Model along with its analysis techniques, an Augmented Fault Tree Analysis, and a Concurrent Event Tree Analysis. The proposed augmented fault tree can identify the potential weak points in software design that may induce unintended software functions or erroneous human procedures. The concurrent event tree can enumerate possible accident sequences due to these weak points

Criticality accident alarm system

International Nuclear Information System (INIS)

Malenfant, R.E.

1991-01-01

The American National Standard ANSI/ANS-8.3-1986, Criticality Accident Alarm System provides guidance for the establishment and maintenance of an alarm system to initiate personnel evacuation in the event of inadvertent criticality. In addition to identifying the physical features of the components of the system, the characteristics of accidents of concern are carefully delineated. Unfortunately, this ANSI Standard has led to considerable confusion in interpretation, and there is evidence that the ''minimum accident of concern'' may not be appropriate. Furthermore, although intended as a guide, the provisions of the standard are being rigorously applied, sometimes with interpretations that are not consistent. Although the standard is clear in the use of absorbed dose in free air of 20 rad, at least one installation has interpreted the requirement to apply to dose in soft tissue. The standard is also clear in specifying the response to both neutrons and gamma rays. An assembly of uranyl fluoride enriched to 5% 235 U was operated to simulate a potential accident. The dose, delivered in a free run excursion 2 m from the surface of the vessel, was greater than 500 rad, without ever exceeding a rate of 20 rad/min, which is the set point for activating an alarm that meets the standard. The presence of an alarm system would not have prevented any of the five major accidents in chemical operations nor is it absolutely certain that the alarms were solely responsible for reducing personnel exposures following the accident. Nevertheless, criticality alarm systems are now the subject of great effort and expense. 13 refs

Case for integral core-disruptive accident analysis

International Nuclear Information System (INIS)

Luck, L.B.; Bell, C.R.

1985-01-01

Integral analysis is an approach used at the Los Alamos National Laboratory to cope with the broad multiplicity of accident paths and complex phenomena that characterize the transition phase of core-disruptive accident progression in a liquid-metal-cooled fast breeder reactor. The approach is based on the combination of a reference calculation, which is intended to represent a band of similar accident paths, and associated system- and separate-effect studies, which are designed to determine the effect of uncertainties. Results are interpreted in the context of a probabilistic framework. The approach was applied successfully in two studies; illustrations from the Clinch River Breeder Reactor licensing assessment are included

Accident analysis. A review of the various accidents classifications

International Nuclear Information System (INIS)

Martin Martin, L.; Figueras, J.M.

1982-01-01

The objective of the accident analysis, in relation with the safety evaluation, environmental impact and emergency planning, should be to identify the total risk to the population and workers from potential accidents in the facility, analizing it over full spectrum of severity. (auth.)

Preliminary H{sub 2} Combustion Analysis in the Containment of APR1400 for SBLOCA Accident using a Multi-Dimensional H{sub 2} Analysis System

Energy Technology Data Exchange (ETDEWEB)

Kang, Hyung Seok; Kim, Jongtae; Kim, Sang-Baik; Hong, Seong-Wan [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

The COM3D analyze an overpressure buildup resulting from a propagation of hydrogen flame along the structure and wall in the containment using the hydrogen distribution result calculated by the GASFLOW. The MAAP evaluates a hydrogen source during a severe accident and transfer it to the GASFLOW. We performed a hydrogen combustion analysis using the multidimensional hydrogen analysis system for a station blackout (SBO) accident under the assumption of 100% metal-water reaction in the reactor vessel. The COM3D results showed that the pressure buildup was about 250 kPa because the flame speed was not increased above 300 m/s and the pressure wave passed through the open spaces in the large containment. To increase the reliability of the COM3D calculation, it is necessary to perform the hydrogen combustion analysis for another accident such as a small break loss of coolant (SBLOCA). KAERI performed a hydrogen combustion analysis for a SBLOCA accident using the multi-dimensional hydrogen analysis system under the assumption of 100% metal-water reaction in the reactor vessel. From the COM3D results, we can know that the pressure buildup was approximately 310 kPa because the flame speed was not increased above 100 m/s owing to the high steam concentration and low oxygen concentration in the hydrogen distributed region of the containment. The predicted maximum overpressure in the SBLOCA accident is similar to that of the COM3D results for the SBO accident. Thus, we found that the maximum overpressure due to the hydrogen combustion in the containment may depend on the amount of hydrogen mass released from the reactor vessel.

The development of severe accident analysis technology

Energy Technology Data Exchange (ETDEWEB)

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

1993-07-01

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

The effect of system modeling on the Fukushima accident evolution

Energy Technology Data Exchange (ETDEWEB)

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

2015-07-01

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

Power Excursion Accident Analysis of Research Water Reactor

International Nuclear Information System (INIS)

Khaled, S.M.; Doaa, G.M.

2009-01-01

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

Neural network-based expert system for severe accident management

International Nuclear Information System (INIS)

Klopp, G.T.; Silverman, E.B.

1992-01-01

This paper presents the results of the second phase of a three-phase Severe Accident Management expert system program underway at Commonwealth Edison Company (CECo). Phase I successfully demonstrated the feasibility of Artificial Neural Networks to support several of the objectives of severe accident management. Simulated accident scenarios were generated by the Modular Accident Analysis Program (MAAP) code currently in use by CECo as part of their Individual Plant Evaluations (IPE)/Accident Management Program. The primary objectives of the second phase were to develop and demonstrate four capabilities of neural networks with respect to nuclear power plant severe accident monitoring and prediction. The results of this work would form the foundation of a demonstration system which included expert system performance features. These capabilities included the ability to: (1) Predict the time available prior to support plate (and reactor vessel) failure; (2) Calculate the time remaining until recovery actions were too late to prevent core damage; (3) Predict future parameter values of each of the MAAP parameter variables; and (4) Detect simulated sensor failure and provide best-value estimates for further processing in the presence of a sensor failure. A variety of accident scenarios for the Zion and Dresden plants were used to train and test the neural network expert system. These included large and small break LOCAs as well as a range of transient events. 3 refs., 1 fig., 1 tab

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jin Yong; Lee, Jeong Hun; Kim, Jong Uk; Yoo, Tae Geun; Chung, Soon Il; Kim, Min Gi [FNC Technology Co., Seoul (Korea, Republic of)

2010-12-15

The objectives of this study are to produce data for level 2 PSA and evaluation results of severe accident by analyzing severe accident sequence of transient events, producing fault tree of containment systems and evaluating direct containment heating of the SMART. In this project, severe accident analysis results were produced for general transient, loss of feedwater, station blackout, and steam line break events, and based on the results, design safety of SMART was verified. Also, direct containment heating phenomenon of the SMART was evaluated using TCE methodology. For level 2 PSA, fault tree of the containment isolation system, reactor cavity flooding system, plant chilled water system, and reactor containment building HVAC system was produced and analyzed

Accident analysis of heat pipe cooled and AMTEC conversion space reactor system

International Nuclear Information System (INIS)

Yuan, Yuan; Shan, Jianqiang; Zhang, Bin; Gou, Junli; Bo, Zhang; Lu, Tianyu; Ge, Li; Yang, Zijiang

2016-01-01

Highlights: • A transient analysis code TAPIRS for HPS has been developed. • Three typical accidents are analyzed using TAPIRS. • The reactor system has the self-stabilization ability under accident conditions. - Abstract: A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of

Nuclear fuel cycle facility accident analysis handbook

International Nuclear Information System (INIS)

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

1988-05-01

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

Environmental decision support system on base of geoinformational technologies for the analysis of nuclear accident consequences

International Nuclear Information System (INIS)

Haas, T.C.; Maigan, M.; Arutyunyan, R.V.; Bolshov, L.A.; Demianov, V.V.

1996-01-01

The report deals with description of the concept and prototype of environmental decision support system (EDSS) for the analysis of late off-site consequences of severe nuclear accidents and analysis, processing and presentation of spatially distributed radioecological data. General description of the available software, use of modem achievements of geostatistics and stochastic simulations for the analysis of spatial data are presented and discussed

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

Waldman, R.M.; Vertullo, A.C.

1987-01-01

The purpose of the present work is the analysis of reactivity transients in MTR cores with LEU and HEU fuels. The analysis includes the following aspects: the phenomenology of the principal events of the accident that takes place, when a reactivity of more than 1$ is inserted in a critical core in less than 1 second. The description of the accident that happened in the RA-2 critical facility in September 1983. The evaluation of the accident from different points of view: a) Theoretical and qualitative analysis; b) Paret Code calculations; c) Comparison with Spert I and Cabri experiments and with post-accident inspections. Differences between LEU and HEU RA-2 cores. (Author)

Severe accident analysis using MARCH 1.0 code

International Nuclear Information System (INIS)

Guimaraes, A.C.F.

1987-09-01

The description and utilization of the MARCH 1.0 computer code, which aim to analyse physical phenomena associated with core meltdown accidents in PWR type reactors, are presented. The primary system is modeled as a single volume which is partitioned into a gas (steam and hydrogen) region and a water region. March predicts blowdown from the primary system in single phase. Based on results of the probabilistic safety analysis for the Zion and Indian Point Nuclear Power Plants, the S 2 HFX sequence accident for Angra-1 reactor is studied. The S 2 HFX sequence means that the loss of coolant accident occurs through small break in primary system with bot total failures of the reactor safety system and containment in yours recirculation modes, leading the core melt and the containment failure due to overpressurization. The obtained results were considered reasonable if compared with the results obtained for the Zion and Indian Point nuclear power plants. (Author) [pt

Analysis of helium purification system capability during water ingress accident in RDE

Science.gov (United States)

Sriyono; Kusmastuti, Rahayu; Bakhri, Syaiful; Sunaryo, Geni Rina

2018-02-01

The water ingress accident caused by steam generator tube rupture (SGTR) in RDE (Experimental Power Reactor) must be anticipated. During the accident, steam from secondary system diffused and mixed with helium gas in the primary coolant. To avoid graphite corrosion in the core, steam will be removed by Helium purification system (HPS). There are two trains in HPS, first train for normal operation and the second for the regeneration and accident. The second train is responsible to clean the coolant during accident condition. The second train is equipped with additional component, i.e. water cooler, post accident blower, and water separator to remove this mixture gas. During water ingress, the water release from rupture tube is mixed with helium gas. The water cooler acts as a steam condenser, where the steam will be separated by water separator from the helium gas. This paper analyses capability of HPS during water ingress accident. The goal of the research is to determine the time consumed by HPS to remove the total amount of water ingress. The method used is modelling and simulation of the HPS by using ChemCAD software. The BDBA and DBA scenarios will be simulated. In BDBA scenario, up to 110 kg of water is assumed to infiltrate to primary coolant while DBA is up to 35 kg. By using ChemCAD simulation, the second train will purify steam ingress maximum in 0.5 hours. The HPS of RDE has a capability to anticipate the water ingress accident.

Status and functioning of the European Commission's major accident reporting system

International Nuclear Information System (INIS)

Kirchsteiger, C.

1999-01-01

This paper describes the background, functioning and status of the European Commission's Major Accident Reporting System (MARS), dedicated to collect, in a consistent way, data on major industrial accidents involving dangerous substances from the Member States of the European Union, to analyse and statistically process them, and to create subsets of all non-confidential accidents data and analysis results for export to all Member States. This modern information exchange and analysis tool is made up of two connected parts: one for each local unit (i.e., for the Competent Authority of each EU Member State), and one central part for the European Commission. The local, as well as the central parts of this information network, can serve both as data logging systems and, on different levels of complexity, as data analysis tools. The central database allows complex cluster and pattern analysis, identifying and analysing the succession of the disruptive factors leading to an accident. On this basis, 'lessons learned' can be formulated for the industry for the purposes of further accident prevention. Further, results from analysing data of major industrial accidents reported to MARS are presented. It can be shown that some of the main assumptions in the new 'Seveso II Directive' can directly be validated from MARS data. (Copyright (c) 1999 Elsevier Science B.V., Amsterdam. All rights reserved.)

[Hospital information system performance for road traffic accidents analysis in a hospital recruitment based area].

Science.gov (United States)

Jannot, A-S; Fauconnier, J

2013-06-01

Road traffic accidents in France are mainly analyzed through reports completed by the security forces (police and gendarmerie). But the hospital information systems can also identify road traffic accidents via specific documentary codes of the International Classification of Diseases (ICD-10). The aim of this study was therefore to determine whether hospital stays consecutive to road traffic accident were truly identified by these documentary codes in a facility that collects data routinely and to study the consistency of results from hospital information systems and from security forces during the 2002-2008 period. We retrieved all patients for whom a documentary code for road traffic accident was entered in 2002-2008. We manually checked the concordance of documentary code for road traffic accident and trauma origin in 350 patient files. The number of accidents in the Grenoble area was then inferred by combining with hospitalization regional data and compared to the number of persons injured by traffic accidents declared by the security force. These hospital information systems successfully report road traffic accidents with 96% sensitivity (95%CI: [92%, 100%]) and 97% specificity (95%CI: [95%, 99%]). The decrease in road traffic accidents observed was significantly less than that observed was significantly lower than that observed in the data from the security force (45% for security force data against 27% for hospital data). Overall, this study shows that hospital information systems are a powerful tool for studying road traffic accidents morbidity in hospital and are complementary to security force data. Copyright © 2013 Elsevier Masson SAS. All rights reserved.

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

Energy Technology Data Exchange (ETDEWEB)

Lee, Young Jin; Chung, Bub Dong [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2004-07-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures.

Coupling the severe accident code SCDAP with the system thermal hydraulic code MARS

International Nuclear Information System (INIS)

Lee, Young Jin; Chung, Bub Dong

2004-01-01

MARS is a best-estimate system thermal hydraulics code with multi-dimensional modeling capability. One of the aims in MARS code development is to make it a multi-functional code system with the analysis capability to cover the entire accident spectrum. For this purpose, MARS code has been coupled with a number of other specialized codes such as CONTEMPT for containment analysis, and MASTER for 3-dimensional kinetics. And in this study, the SCDAP code has been coupled with MARS to endow the MARS code system with severe accident analysis capability. With the SCDAP, MARS code system now has acquired the capability to simulate such severe accident related phenomena as cladding oxidation, melting and slumping of fuel and reactor structures

A database system for the management of severe accident risk information, SARD

International Nuclear Information System (INIS)

Ahn, K. I.; Kim, D. H.

2003-01-01

The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies

A database system for the management of severe accident risk information, SARD

Energy Technology Data Exchange (ETDEWEB)

Ahn, K. I.; Kim, D. H. [KAERI, Taejon (Korea, Republic of)

2003-10-01

The purpose of this paper is to introduce main features and functions of a PC Windows-based database management system, SARD, which has been developed at Korea Atomic Energy Research Institute for automatic management and search of the severe accident risk information. Main functions of the present database system are implemented by three closely related, but distinctive modules: (1) fixing of an initial environment for data storage and retrieval, (2) automatic loading and management of accident information, and (3) automatic search and retrieval of accident information. For this, the present database system manipulates various form of the plant-specific severe accident risk information, such as dominant severe accident sequences identified from the plant-specific Level 2 Probabilistic Safety Assessment (PSA) and accident sequence-specific information obtained from the representative severe accident codes (e.g., base case and sensitivity analysis results, and summary for key plant responses). The present database system makes it possible to implement fast prediction and intelligent retrieval of the required severe accident risk information for various accident sequences, and in turn it can be used for the support of the Level 2 PSA of similar plants and for the development of plant-specific severe accident management strategies.

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Detection device for off-gas system accidents

International Nuclear Information System (INIS)

Kubota, Ryuji; Tsuruoka, Ryozo; Yamanari, Shozo.

1984-01-01

Purpose: To rapidly isolate the off-gas system by detecting the off-gas system failure accident in a short time. Constitution: Radiation monitors are disposed to ducts connecting an exhaust gas area and an air conditioning system as a portion of a turbine building. The ducts are disposed independently such that they ventilate only the atmosphere in the exhaust gas area and do not mix the atmosphere in the turbine building. Since radioactivity issued upon off-gas accidents to the exhaust gas area is sucked to the duct, it can be detected by radiation detection monitors in a short time after the accident. Further, since the operator judges it as the off-gas system accident, the off-gas system can be isolated in a short time after the accident. (Moriyama, K.)

Computerized accident management support system: development for severe accident management

International Nuclear Information System (INIS)

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

1998-01-01

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

A systemic analysis of patterns of organizational breakdowns in accidents: A case from Helicopter Emergency Medical Service (HEMS) operations

International Nuclear Information System (INIS)

Kontogiannis, Tom; Malakis, Stathis

2012-01-01

In recent years, many accident models and techniques have shifted their focus from shortfalls in the actions of practitioners to systemic causes in the organization. Accident investigation techniques (e.g., STAMP) have been developed that looked into the flaws of control processes in the organization. Organizational models have looked into general patterns of breakdown related to structural vulnerabilities and gradual degradation of performance. Although some degree of cross-fertilization has been developed between these two trends, safety analysts are left on their own to integrate this gap between control flaws and patterns of organizational breakdown in accident investigation. This article attempts to elaborate the control dynamics of the Systems Theoretic Accident Model and Process (STAMP) technique on the basis of a theoretical model of organizational viability (i.e., the Viable Systems Model). The joint STAMP–VSM framework is applied to an accident from a Helicopter Emergency Medical Service (HEMS) organization to help analysts progress from the analysis of control flaws to the underlying patterns of breakdown. The joint framework may help analysts to rethink the safety organization, model new information loops and constraints, look at the adaptation and steering functions of the organization and finally, develop high leverage interventions. - Highlights: ► This article bridges the gap between two parallel trends in systemic accident models. ► Investigation techniques (i.e., STAMP) have looked into the flaws of safety management processes. ► The literature has highlighted many patterns (or archetypes) of organizational breakdowns. ► The Viable System Model is used with STAMP to link control flaws and organizational breakdowns.

APR1400 CEA Withdrawal at Power Accident Analysis using KNAP

International Nuclear Information System (INIS)

Lee, Dong-Hyuk; Yang, Chang-Keun; Kim, Yo-Han; Sung, Chang-Kyung

2006-01-01

KEPRI (Korea Electric Power Research Institute) has been developing safety analysis methodology for non- LOCA (Loss Of Coolant Accident) analysis of OPR1000 (Optimized Power Reactor 1000, formerly KSNP). The new methodology, named KNAP (Korea Non-LOCA Analysis Package), uses RETRAN as the main system analysis code. RETRAN code is a non- LOCA safety analysis code developed by EPRI. The new methodology will replace existing CE (Combustion Engineering) supplied codes and methodologies currently used in non-LOCA analysis of OPR1000. In this paper, we apply KNAP methodology to APR1400 (Advanced Power Reactor 1400). The CEA (Control Element Assembly) withdrawal at power accident is one of the 'reactivity and power distribution anomalies' events and the results are typically described in the chapter 15.4.2 of SAR (Safety Analysis Report). The APR1400 has been designed to generate 1,400MWe of electricity with advanced features for greatly enhanced safety and economic goals. The CEA withdrawal at power analysis in APR1400 SSAR (Standard Safety Analysis Report) is analyzed with CESEC-III computer code. In this study, to confirm the applicability of the KNAP methodology and code system to APR1400, CEA withdrawal at power accident is analyzed using RETRAN code and it is compared with results from APR1400 SSAR

RASCAL [Radiological Assessment System for Consequence AnaLysis]: A screening model for estimating doses from radiological accidents

International Nuclear Information System (INIS)

Sjoreen, A.L.; Athey, G.F.; Sakenas, C.A.; McKenna, T.J.

1988-01-01

The Radiological Assessment System for Consequence AnaLysis (RASCAL) is a new MS-DOS-based dose assessment model which has been written for the US Nuclear Regulatory Commission for use during response to radiological emergencies. RASCAL is designed to provide crude estimates of the effects of an accident while the accident is in progress and only limited information is available. It has been designed to be very simple to use and to run quickly. RASCAL is unique in that it estimates the source term based on fundamental plant conditions and does not rely solely on release rate estimation (e.g., Ci/sec of I-131). Therefore, it can estimate consequences of accidents involving unmonitored pathways or projected failures. RASCAL will replace the older model, IRDAM. 6 refs

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

Human error and the problem of causality in analysis of accidents

DEFF Research Database (Denmark)

Rasmussen, Jens

1990-01-01

, designers or managers have played a major role. There are, however, several basic problems in analysis of accidents and identification of human error. This paper addresses the nature of causal explanations and the ambiguity of the rules applied for identification of the events to include in analysis......Present technology is characterized by complexity, rapid change and growing size of technical systems. This has caused increasing concern with the human involvement in system safety. Analyses of the major accidents during recent decades have concluded that human errors on part of operators...

A study on risk analysis for loading and un-loading accident

International Nuclear Information System (INIS)

Watabe, N.; Suzuki, H.; Saegusa, T.

1998-01-01

Low Level Waste packages are transported from each Japanese nuclear power plants to Rokkasho-Mura by exclusive ship. These packages are contained in half-height 5 ton containers. The handling system for loading and unloading containers is composed of the 25 ton crane, the cell-guide system and transport trucks. These systems are mostly automated and under computer control. By design, the whole handling system should be highly protected from any accident. However unknown causes for accidents might be concealed in this handling system, because of complicated system interaction between computer control and human operation. The representative 25 ton bridge type crane was analyzed in this assessment. As the first step, causes of drop accidents were analyzed using design drawing of the crane and its system operation flow chart as inputs to the analysis. After analysis the protection methods were reviewed, and where necessary, revised in each step accident cause. Those results were rearranged by fault trees for each cause. To provide quantitative details of operational interactions, crane operators and safety supervisors were consulted. Based on their experience, a method to determine probabilities of basic events was tentatively adopted. According to this assessment, each protection method was clarified and some weak points of the loading and un-loading process were able to be identified. Figure 1 shows schematically the sequential steps in the method. As a result of this assessment, the PSA method (including fault trees, etc) was found to be adaptable for the loading and un-loading process (i.e. handling system) and to be effective in understanding the system characteristics. Further, using this PSA analysis method allows transport companies to review protection methods with 'Cost and Benefit' analysis concepts. (authors)

An Evidential Reasoning-Based CREAM to Human Reliability Analysis in Maritime Accident Process.

Science.gov (United States)

Wu, Bing; Yan, Xinping; Wang, Yang; Soares, C Guedes

2017-10-01

This article proposes a modified cognitive reliability and error analysis method (CREAM) for estimating the human error probability in the maritime accident process on the basis of an evidential reasoning approach. This modified CREAM is developed to precisely quantify the linguistic variables of the common performance conditions and to overcome the problem of ignoring the uncertainty caused by incomplete information in the existing CREAM models. Moreover, this article views maritime accident development from the sequential perspective, where a scenario- and barrier-based framework is proposed to describe the maritime accident process. This evidential reasoning-based CREAM approach together with the proposed accident development framework are applied to human reliability analysis of a ship capsizing accident. It will facilitate subjective human reliability analysis in different engineering systems where uncertainty exists in practice. © 2017 Society for Risk Analysis.

Aircraft accident analysis for emergency planning and safety analysis

International Nuclear Information System (INIS)

Nicolosi, S.L.; Jordan, H.; Foti, D.; Mancuso, J.

1996-01-01

Potential aircraft accidents involving facilities at the Rocky Flats Environmental Technology Site (Site) are evaluated to assess their safety significance. This study addresses the probability and facility penetrability of aircraft accidents at the Site. The types of aircraft (large, small, etc.) that may credibly impact the Site determine the types of facilities that may be breached. The methodology used in this analysis follows elements of the draft Department of Energy Standard ''Accident Analysis for Aircraft Crash into Hazardous Facilities'' (July 1995). Key elements used are: the four-factor frequency equation for aircraft accidents; the distance criteria for consideration of airports, airways, and jet routes; the consideration of different types of aircraft; and the Modified National Defense Research Committee (NDRC) formula for projectile penetration, perforation, and minimum resistant thickness. The potential aircraft accident frequency for each type of aircraft applicable to the Site is estimated using a four-factor formula described in the draft Standard. The accident frequency is the product of the annual number of operations, probability of an accident, probability density function, and area. The annual number of operations is developed from site-specific and state-wide data

NASA Accident Precursor Analysis Handbook, Version 1.0

Science.gov (United States)

Groen, Frank; Everett, Chris; Hall, Anthony; Insley, Scott

2011-01-01

Catastrophic accidents are usually preceded by precursory events that, although observable, are not recognized as harbingers of a tragedy until after the fact. In the nuclear industry, the Three Mile Island accident was preceded by at least two events portending the potential for severe consequences from an underappreciated causal mechanism. Anomalies whose failure mechanisms were integral to the losses of Space Transportation Systems (STS) Challenger and Columbia had been occurring within the STS fleet prior to those accidents. Both the Rogers Commission Report and the Columbia Accident Investigation Board report found that processes in place at the time did not respond to the prior anomalies in a way that shed light on their true risk implications. This includes the concern that, in the words of the NASA Aerospace Safety Advisory Panel (ASAP), "no process addresses the need to update a hazard analysis when anomalies occur" At a broader level, the ASAP noted in 2007 that NASA "could better gauge the likelihood of losses by developing leading indicators, rather than continue to depend on lagging indicators". These observations suggest a need to revalidate prior assumptions and conclusions of existing safety (and reliability) analyses, as well as to consider the potential for previously unrecognized accident scenarios, when unexpected or otherwise undesired behaviors of the system are observed. This need is also discussed in NASA's system safety handbook, which advocates a view of safety assurance as driving a program to take steps that are necessary to establish and maintain a valid and credible argument for the safety of its missions. It is the premise of this handbook that making cases for safety more experience-based allows NASA to be better informed about the safety performance of its systems, and will ultimately help it to manage safety in a more effective manner. The APA process described in this handbook provides a systematic means of analyzing candidate

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

International Nuclear Information System (INIS)

Castro, A.; Buedo, J.L.; Borondo, L.; Lopez, N.

2001-01-01

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

Accident Analysis and Highway Safety

Directory of Open Access Journals (Sweden)

Omar Noorliyana

2017-01-01

Full Text Available Since 2010, Federal Route FT050 (Jalan Batu Pahat-Kluang has undergone many changes, including the improvement of geometric features (i.e., construction of median, dedicated U-turns and additional lanes and upgrading the quality of the road surface. Unfortunately, even with these enhancements, accidents continue to occur along this route. This study covered both accident analysis and blackspot study. Accident point weightage was used to identify blackspot locations. The results reveal hazardous road locations and blackspot ranking along the route.

Analysis of loss of coolant accident and emergency core cooling system

International Nuclear Information System (INIS)

Abe, Kiyoharu; Kobayashi, Kenji; Hayata, Kunihisa; Tasaka, Kanji; Shiba, Masayoshi

1977-01-01

In this paper, the analysis for the performance evaluation of emergency core cooling system is described, which is the safety protection device to the loss of coolant accidents due to the break of primary cooling pipings of light water reactors. In the LOCA analysis for the performance evaluation of ECCS, it must be shown that a reactor core keeps the form which can be cooled with the ECCS in case of LOCA, and the overheat of the core can be prevented. Namely, the shattering of fuel cladding tubes is never to occur, and for the purpose, the maximum temperature of Zircaloy 2 or 4 cladding tubes must be limited to 1200 deg C, and the relative thickness of oxide film must be below 15%. The calculation for determining the temperature of cladding tubes in case of the LOCA in BWRs and PWRs is explained. First, the primary cooling system, the ECCS and the related installations of BWRs and PWRs are outlined. The code systems for LOCA/ECCS analysis are divid ed into several steps, such as blowdown process, reflooding process and heatup calculation. The examples of the sensitivity analysis of the codes are shown. The LOCA experiments carried out so far in Japan and foreign countries and the LOCA analysis of a BWR with RELAP-4J code are described. The guidance for the performance evaluation of ECCS was established in 1975 by the Reactor Safety Deliberation Committee in Japan, and the contents are quoted. (Kako, I.)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Accident analysis for PRC-II reactor

International Nuclear Information System (INIS)

Wei Yongren; Tang Gang; Wu Qing; Lu Yili; Liu Zhifeng

1997-12-01

The computer codes, calculation models, transient results, sensitivity research, design improvement, and safety evaluation used in accident analysis for PRC-II Reactor (The Second Pulsed Reactor in China) are introduced. PRC-II Reactor is built in big populous city, so the public pay close attention to reactor safety. Consequently, Some hypothetical accidents are analyzed. They include an uncontrolled control rod withdrawal at rated power, a pulse rod ejection at rated power, and loss of coolant accident. Calculation model which completely depict the principle and process for each accident is established and the relevant analysis code is developed. This work also includes comprehensive computing and analyzing transients for each accident of PRC-II Reactor; the influences in the reactor safety of all kind of sensitive parameters; evaluating the function of engineered safety feature. The measures to alleviate the consequence of accident are suggested and taken in the construction design of PRC-II Reactor. The properties of reactor safety are comprehensively evaluated. A new advanced calculation model (True Core Uncovered Model) of LOCA of PRC-II Reactor and the relevant code (MCRLOCA) are first put forward

The accident analysis of mobile mine machinery in Indian opencast coal mines.

Science.gov (United States)

Kumar, R; Ghosh, A K

2014-01-01

This paper presents the analysis of large mining machinery related accidents in Indian opencast coal mines. The trends of coal production, share of mining methods in production, machinery deployment in open cast mines, size and population of machinery, accidents due to machinery, types and causes of accidents have been analysed from the year 1995 to 2008. The scrutiny of accidents during this period reveals that most of the responsible factors are machine reversal, haul road design, human fault, operator's fault, machine fault, visibility and dump design. Considering the types of machines, namely, dumpers, excavators, dozers and loaders together the maximum number of fatal accidents has been caused by operator's faults and human faults jointly during the period from 1995 to 2008. The novel finding of this analysis is that large machines with state-of-the-art safety system did not reduce the fatal accidents in Indian opencast coal mines.

Cost-effectiveness analysis of countermeasures using accident consequence assessment models

International Nuclear Information System (INIS)

Alonso, A.; Gallego, E.

1987-01-01

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

The causing model of accidents and preventing system of small mines

Energy Technology Data Exchange (ETDEWEB)

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

2008-06-15

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

[Violence and accidents among older and younger adults: evidence from the Surveillance System for Violence and Accidents (VIVA), Brazil].

Science.gov (United States)

Luz, Tatiana Chama Borges; Malta, Deborah Carvalho; Sá, Naíza Nayla Bandeira de; Silva, Marta Maria Alves da; Lima-Costa, Maria Fernanda

2011-11-01

Data from the Brazilian Surveillance System for Violence and Accidents (VIVA) in 2009 were used to examine socio-demographic characteristics, outcomes, and types of accidents and violence treated at 74 sentinel emergency services in 23 Brazilian State capitals and the Federal District. The analysis included 25,201 individuals aged > 20 years (10.1% > 60 years); 89.3% were victims of accidents and 11.9% victims of violence. Hospitalization was the outcome in 11.1% of cases. Compared to the general population, there were more men and non-white individuals among victims of accidents, and especially among victims of violence. As compared to younger adults (20-59 years), accidents and violence against elderly victims showed less association with alcohol, a higher proportion of domestic incidents, more falls and pedestrian accidents, and aggression by family members. Policies for the prevention of accidents and violence should consider the characteristics of these events in the older population.

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

International Nuclear Information System (INIS)

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

2015-01-01

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

Bibliography for nuclear criticality accident experience, alarm systems, and emergency management

International Nuclear Information System (INIS)

Putman, V.L.

1995-09-01

The characteristics, detection, and emergency management of nuclear criticality accidents outside reactors has been an important component of criticality safety for as long as the need for this specialized safety discipline has been recognized. The general interest and importance of such topics receives special emphasis because of the potentially lethal, albeit highly localized, effects of criticality accidents and because of heightened public and regulatory concerns for any undesirable event in nuclear and radiological fields. This bibliography lists references which are potentially applicable to or interesting for criticality alarm, detection, and warning systems; criticality accident emergency management; and their associated programs. The lists are annotated to assist bibliography users in identifying applicable: industry and regulatory guidance and requirements, with historical development information and comments; criticality accident characteristics, consequences, experiences, and responses; hazard-, risk-, or safety-analysis criteria; CAS design and qualification criteria; CAS calibration, maintenance, repair, and testing criteria; experiences of CAS designers and maintainers; criticality accident emergency management (planning, preparedness, response, and recovery) requirements and guidance; criticality accident emergency management experience, plans, and techniques; methods and tools for analysis; and additional bibliographies

A methodology for radiological accidents analysis in industrial gamma radiography

International Nuclear Information System (INIS)

Silva, F.C.A. da.

1990-01-01

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

Analysis of traffic accidents in Romania, 2009.

Science.gov (United States)

Călinoiu, Geovana; Minca, Dana Galieta; Furtunescu, Florentina Ligia

2012-01-01

This paper aimed to underline the main consequences of traffic accidents in Romania 2009 and their associated causes or circumstances. We identified some problematic geographic areas, some critical months or moments of the day and also the most frequent causes; all these should become targets for the future planning. The current analysis provides some priority criteria for public health interventions. So, the future national road safety strategy should be in line with the EU objectives, but also with the national priorities. Romania is far away from the average EU target for 2010 of halving the death by traffic accidents registered in 2001. To describe the circumstances and the consequences related to traffic accidents registered in Romania, for the year 2009. An ecological study was conducted. The traffic accidents circumstances were analyzed in terms of magnitude, geographic space, time and cause. The consequences were analyzed as affected people and damaged cars. A total of 28,627 traffic accidents were registered in Romania during the year 2009. 2,796 people were killed and 27,968 were hospitalized and 42,443 cars were damaged. 3 of 4 accidents were caused by violations on behalf of the car drivers. Most common violations in car drivers were excess of speed and priority violations (52.4%). Among the pedestrians, 7 of 10 accidents were caused by illegal crossing. A higher number of accidents occurred during the summer months and during the evening hours (from 5.00 pm till 8.00 pm). The traffic accidents represent a real public health problem in Romania and a serious burden for the health system. The gap between Romania and the other EU member states needs to be diminished in the next decade. In this purpose, the future national road safety strategy should be in line with the EU objectives, but also with the national priorities. Research is needed to understand the causes and the socio-economical impact of traffic accidents and to define appropriate national

Probabilistic Dynamics for Integrated Analysis of Accident Sequences considering Uncertain Events

Directory of Open Access Journals (Sweden)

Robertas Alzbutas

2015-01-01

Full Text Available The analytical/deterministic modelling and simulation/probabilistic methods are used separately as a rule in order to analyse the physical processes and random or uncertain events. However, in the currently used probabilistic safety assessment this is an issue. The lack of treatment of dynamic interactions between the physical processes on one hand and random events on the other hand causes the limited assessment. In general, there are a lot of mathematical modelling theories, which can be used separately or integrated in order to extend possibilities of modelling and analysis. The Theory of Probabilistic Dynamics (TPD and its augmented version based on the concept of stimulus and delay are introduced for the dynamic reliability modelling and the simulation of accidents in hybrid (continuous-discrete systems considering uncertain events. An approach of non-Markovian simulation and uncertainty analysis is discussed in order to adapt the Stimulus-Driven TPD for practical applications. The developed approach and related methods are used as a basis for a test case simulation in view of various methods applications for severe accident scenario simulation and uncertainty analysis. For this and for wider analysis of accident sequences the initial test case specification is then extended and discussed. Finally, it is concluded that enhancing the modelling of stimulated dynamics with uncertainty and sensitivity analysis allows the detailed simulation of complex system characteristics and representation of their uncertainty. The developed approach of accident modelling and analysis can be efficiently used to estimate the reliability of hybrid systems and at the same time to analyze and possibly decrease the uncertainty of this estimate.

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

International Nuclear Information System (INIS)

Farmer, Mitchell T.; Bunt, R.; Corradini, M.; Ellison, Paul B.; Francis, M.; Gabor, John D.; Gauntt, R.; Henry, C.; Linthicum, R.; Luangdilok, W.; Lutz, R.; Paik, C.; Plys, M.; Rabiti, Cristian; Rempe, J.; Robb, K.; Wachowiak, R.

2015-01-01

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy's (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

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

Energy Technology Data Exchange (ETDEWEB)

Farmer, Mitchell T. [Argonne National Lab. (ANL), Argonne, IL (United States); Bunt, R. [Southern Nuclear, Atlanta, GA (United States); Corradini, M. [Univ. of Wisconsin, Madison, WI (United States); Ellison, Paul B. [GE Power and Water, Duluth, GA (United States); Francis, M. [Argonne National Lab. (ANL), Argonne, IL (United States); Gabor, John D. [Erin Engineering, Walnut Creek, CA (United States); Gauntt, R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Henry, C. [Fauske and Associates, Burr Ridge, IL (United States); Linthicum, R. [Exelon Corp., Chicago, IL (United States); Luangdilok, W. [Fauske and Associates, Burr Ridge, IL (United States); Lutz, R. [PWR Owners Group (PWROG); Paik, C. [Fauske and Associates, Burr Ridge, IL (United States); Plys, M. [Fauske and Associates, Burr Ridge, IL (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rempe, J. [Rempe and Associates LLC, Idaho Falls, ID (United States); Robb, K. [Argonne National Lab. (ANL), Argonne, IL (United States); Wachowiak, R. [Electric Power Research Inst. (EPRI), Knovville, TN (United States)

2015-01-31

The overall objective of this study was to conduct a technology gap evaluation on accident tolerant components and severe accident analysis methodologies with the goal of identifying any data and/or knowledge gaps that may exist, given the current state of light water reactor (LWR) severe accident research, and additionally augmented by insights obtained from the Fukushima accident. The ultimate benefit of this activity is that the results can be used to refine the Department of Energy’s (DOE) Reactor Safety Technology (RST) research and development (R&D) program plan to address key knowledge gaps in severe accident phenomena and analyses that affect reactor safety and that are not currently being addressed by the industry or the Nuclear Regulatory Commission (NRC).

Loss-of-coolant accident analysis of the Savannah River new production reactor design

International Nuclear Information System (INIS)

Maloney, K.J.; Pryor, R.J.

1990-11-01

This document contains the loss-of-coolant accident analysis of the representative design for the Savannah River heavy water new production reactor. Included in this document are descriptions of the primary system, reactor vessel, and loss-of-coolant accident computer input models, the results of the cold leg and hot leg loss-of-coolant accident analyses, and the results of sensitivity calculations for the cold leg loss-of-coolant accident. 5 refs., 50 figs., 4 tabs

Historical analysis of US pipeline accidents triggered by natural hazards

Science.gov (United States)

Girgin, Serkan; Krausmann, Elisabeth

2015-04-01

Natural hazards, such as earthquakes, floods, landslides, or lightning, can initiate accidents in oil and gas pipelines with potentially major consequences on the population or the environment due to toxic releases, fires and explosions. Accidents of this type are also referred to as Natech events. Many major accidents highlight the risk associated with natural-hazard impact on pipelines transporting dangerous substances. For instance, in the USA in 1994, flooding of the San Jacinto River caused the rupture of 8 and the undermining of 29 pipelines by the floodwaters. About 5.5 million litres of petroleum and related products were spilled into the river and ignited. As a results, 547 people were injured and significant environmental damage occurred. Post-incident analysis is a valuable tool for better understanding the causes, dynamics and impacts of pipeline Natech accidents in support of future accident prevention and mitigation. Therefore, data on onshore hazardous-liquid pipeline accidents collected by the US Pipeline and Hazardous Materials Safety Administration (PHMSA) was analysed. For this purpose, a database-driven incident data analysis system was developed to aid the rapid review and categorization of PHMSA incident reports. Using an automated data-mining process followed by a peer review of the incident records and supported by natural hazard databases and external information sources, the pipeline Natechs were identified. As a by-product of the data-collection process, the database now includes over 800,000 incidents from all causes in industrial and transportation activities, which are automatically classified in the same way as the PHMSA record. This presentation describes the data collection and reviewing steps conducted during the study, provides information on the developed database and data analysis tools, and reports the findings of a statistical analysis of the identified hazardous liquid pipeline incidents in terms of accident dynamics and

EUROCONTROL-Systemic Occurrence Analysis Methodology (SOAM)-A 'Reason'-based organisational methodology for analysing incidents and accidents

International Nuclear Information System (INIS)

Licu, Tony; Cioran, Florin; Hayward, Brent; Lowe, Andrew

2007-01-01

The Safety Occurrence Analysis Methodology (SOAM) developed for EUROCONTROL is an accident investigation methodology based on the Reason Model of organisational accidents. The purpose of a SOAM is to broaden the focus of an investigation from human involvement issues, also known as 'active failures of operational personnel' under Reason's original model, to include analysis of the latent conditions deeper within the organisation that set the context for the event. Such an approach is consistent with the tenets of Just Culture in which people are encouraged to provide full and open information about how incidents occurred, and are not penalised for errors. A truly systemic approach is not simply a means of transferring responsibility for a safety occurrence from front-line employees to senior managers. A consistent philosophy must be applied, where the investigation process seeks to correct deficiencies wherever they may be found, without attempting to apportion blame or liability

Accident consequence calculations for project W-058 safety analysis

International Nuclear Information System (INIS)

Van Keuren, J.C.

1997-01-01

This document describes the calculations performed to determine the accident consequences for the W-058 safety analysis. Project W-058 is the replacement cross site transfer system (RCSTS), which is designed to transort liquid waste between the 200 W and 200 E areas. Calculations for RCSTS safety analyses used the same methods as the calculations for the Tank Waste Remediation System (TWRS) Basis for Interim Operation (BIO) and its supporting calculation notes. Revised analyses were performed for the spray and pool leak accidents since the RCSTS flows and pressures differ from those assumed in the TWRS BIO. Revision 1 of the document incorporates review comments

The role of systems availability and operator actions in accident management

International Nuclear Information System (INIS)

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

1988-01-01

Traditional analyses of severe accidents, such as those presented in Probabilistic Risk Assessment (PRA) studies of nuclear power stations, have generally been performed on the assumption that all means of cooling the reactor core are lost and that no operator actions to mitigate the consequences or progression of the severe accident are performed. The assumption to neglect the availability of safety systems and operator actions which do not prevent core melting can lead to erroneous conclusions regarding the plant severe accident profile. Recent work in severe accident management has identified the need to perform analyses which consider all systems availabilities and operator actions, irrespective of their contribution to the prevention of core melting. These new analyses have far reaching conclusions. The analysis results indicate an unacceptably high degree of simplicity in the present severe accident analyses for Probabilistic Risk Assessment studies; the simplicity is in the assumption that systems availabilities and operator actions which do not impact core melt frequency can be neglected in the severe accident analyses. This results in overly pessimistic predictions of the time of core melting and the subsequent potential for recovery of core cooling prior to core melting. This simplicity can have a considerable impact on severe accident decision making, particularly in the evaluation of alternate plant design features and the priorities for research studies

Steady-state thermal hydraulic analysis and flow channel blockage accident analysis of JRR-3 silicide core

International Nuclear Information System (INIS)

Kaminaga, Masanori

1997-03-01

JRR-3 is a light water moderated and cooled, beryllium and heavy water reflected pool type research reactor using low enriched uranium (LEU) plate-type fuels. Its thermal power is 20 MW. The core conversion program from uranium-aluminum (UAl x -Al) dispersion type fuel (aluminide fuel) to uranium-silicon-aluminum (U 3 Si 2 -Al) dispersion type fuel (silicide fuel) is currently conducted at the JRR-3. This report describes about the steady-state thermal hydraulic analysis results and the flow channel blockage accident analysis result. In JRR-3, there are two operation mode. One is high power operation mode up to 20 MW, under forced convection cooling using the primary and the secondary cooling systems. The other is low power operation mode up to 200 kW, under natural circulation cooling between the reactor core and the reactor pool without the primary and the secondary cooling systems. For the analysis of the flow channel blockage accident, COOLOD code was used. On the other hand, steady-state thermal hydraulic analysis for both of the high power operation mode under forced convection cooling and low power operation under natural convection cooling, COOLOD-N2 code was used. From steady-state thermal hydraulic analysis results of both forced and natural convection cooling, fuel temperature, minimum DNBR etc. meet the design criteria and JRR-3 LEU silicide core has enough safety margin under normal operation conditions. Furthermore, flow channel blockage accident analysis results show that one channel flow blockage accident meet the safety criteria for accident conditions which have been established for JRR-3 LEU silicide core. (author)

Risk Analysis of Fukushima Accident using MACCS2

Energy Technology Data Exchange (ETDEWEB)

Lee, Seunghee; Kim, Juyoul; Kim, Sukhoon; Kim, Juyub [FNC Technology Co. Ltd., Yongin (Korea, Republic of)

2014-05-15

It has been three years since Fukushima Daiichi accident had occurred. Many efforts have been done for a restoration, however, radioactive materials are still released resulting in a crucial additional damage to a human health and economics and the scale of damage is not much evaluated. Therefore, an estimation of damage degree caused by the released radioactive materials right after a nuclear accident is essential to cope with additional radioactive problems. Here, we report the risk analysis of Fukushima Dai-ichi accident using MELCOR Accident Consequence Code System 2 (MACCS2), which is the Nuclear Regulatory Commission's (NRC's) code for evaluating off-site consequences. It is used in level-3 Probabilistic Risk Analyses (PRA), for planning purposes, for cost-benefit analyses and so on. The purpose of this study is to estimate radiological doses and health risks of Fukushima Daiichi accident through short- and long-term of lifetime using MACCS2. In summary, the health risk for inhabitants near Fukushima Daiichi NPP has been evaluated by considering the long term radiation effect using MACCS2 code. The result indicates that the occurrence and death rate of the cancer have been increased by the radioactive materials released from Fukushima Daiichi accident. The result obtained in this study may provide new insights for taking action after the nuclear reactor accident to mitigate the released radioactive materials and to prepare the countermeasure.

CANDU severe accident analysis

International Nuclear Information System (INIS)

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

2007-01-01

Romania is a EU member since January first 2007. This country faces now new challenges which imply also the nuclear power reactors now in operation. Romania operates since 1996 a CANDU nuclear power reactor and soon will start up a second unit. In EU PWR reactors are mostly operated, so that the Romania's reactors have to meet EU standards. Safety analysis guidelines require to model severe accidents for reactors of this type. Starting from previous studies a thermal-hydraulic model for a degraded CANDU core was developed. The initiating event is assumed to be a LOCA with simultaneous loss of moderator and coolant and the failure 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 temperatures 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 water tank surrounding the calandria vessel. The thermal hydraulics phenomena described above are modeled, analyzed and compared with the existing data. (authors)

Methodology of a PWR containment analysis during a thermal-hydraulic accident

Energy Technology Data Exchange (ETDEWEB)

Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: dayane.silva@usp.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2015-07-01

The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

Methodology of a PWR containment analysis during a thermal-hydraulic accident

International Nuclear Information System (INIS)

Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S.

2015-01-01

The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code. (author)

Human factors review for Severe Accident Sequence Analysis (SASA)

International Nuclear Information System (INIS)

Krois, P.A.; Haas, P.M.; Manning, J.J.; Bovell, C.R.

1984-01-01

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

Application of forensic image analysis in accident investigations.

Science.gov (United States)

Verolme, Ellen; Mieremet, Arjan

2017-09-01

Forensic investigations are primarily meant to obtain objective answers that can be used for criminal prosecution. Accident analyses are usually performed to learn from incidents and to prevent similar events from occurring in the future. Although the primary goal may be different, the steps in which information is gathered, interpreted and weighed are similar in both types of investigations, implying that forensic techniques can be of use in accident investigations as well. The use in accident investigations usually means that more information can be obtained from the available information than when used in criminal investigations, since the latter require a higher evidence level. In this paper, we demonstrate the applicability of forensic techniques for accident investigations by presenting a number of cases from one specific field of expertise: image analysis. With the rapid spread of digital devices and new media, a wealth of image material and other digital information has become available for accident investigators. We show that much information can be distilled from footage by using forensic image analysis techniques. These applications show that image analysis provides information that is crucial for obtaining the sequence of events and the two- and three-dimensional geometry of an accident. Since accident investigation focuses primarily on learning from accidents and prevention of future accidents, and less on the blame that is crucial for criminal investigations, the field of application of these forensic tools may be broader than would be the case in purely legal sense. This is an important notion for future accident investigations. Copyright © 2017 Elsevier B.V. All rights reserved.

Development of Severe Accident Containment Analysis Package

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

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

Strategy generator in computerized accident management support system

International Nuclear Information System (INIS)

Sirola, M.

1994-02-01

An increased interest for research in the field of accident management of nuclear power plants can be noted. Several international programmes have been started in order to be able to understand the basic physical and chemical phenomena in accident conditions. A feasibility study has shown that it would be possible to design and develop a computerized support system for plant staff in accident situations. To achieve this goal the Halden Project has initiated a research programme on Computerized Accident Management Support (CAMS project). The aim is to utilize the capabilities of computerized tools to support the plant staff during the various accident stages. The system will include identification of the accident state, assessment of the future development of the accident and planning of accident mitigation strategies. A prototype is developed to support operators and the Technical Support Centre in decision making during serious accidents in nuclear power plants. A rule based system has been built to take care of the strategy generation. This system assists plant personnel in planning control proposals and mitigation strategies from normal operation to severe accident conditions. The idea of a safety objective tree and knowledge from the emergency procedures have been used. Future prediction requires good state identification of the plant status and some knowledge about the history of some critical variables. The information needs to be validated as well. Accurate calculations in simulators and a large database including all important information from the plant will help the strategy planning. (orig.). (40 refs., 20 figs.)

Analysis and research status of severe core damage accidents

International Nuclear Information System (INIS)

1984-03-01

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

MELCOR analysis of the TMI-2 accident

International Nuclear Information System (INIS)

Boucheron, E.A.

1990-01-01

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

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

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

Causative Chain Difference for each Type of Accidents in Japanese Maritime Traffic Systems (MTS

Directory of Open Access Journals (Sweden)

Wanginingastuti Mutmainnah

2017-09-01

Full Text Available Causative chain (CC is a failure chain that cause accident as an outcome product of the second step of MOP model, namely line relation analysis (LRA. This CC is a connection of several causative factors (CF, an outcome product of first step of MOP model, namely corner analysis (CA. MOP Model is an abbreviation from 4M Overturned Pyramid, created by authors by combining 2 accident analysis models. There are two steps in this model, namely CA and LRA. Utilizing this model can know what is CF that happen dominantly to the accidents and what is a danger CC that characterize accidents in a certain place and certain period. By knowing the characteristics, the preventive action can be decided to decrease the number of accident in the next period. The aim of this paper is providing the development of MOP Model that has been upgraded and understanding the characteristics of each type accident. The data that is analyzed in this paper is Japanese accidents from 2008 until 2013, which is available on Japan Transportation Safety Board (JTSB’s website. The analysis shows that every type of accidents has a unique characteristic, shown by their CFs and CCs. However, Man Factor is still playing role to the system dominantly.

Development of Database for Accident Analysis in Indian Mines

Science.gov (United States)

Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

2016-10-01

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

K Basins floor sludge retrieval system knockout pot basket fuel burn accident

International Nuclear Information System (INIS)

HUNT, J.W.

1998-01-01

The K Basins Sludge Retrieval System Preliminary Hazard Analysis Report (HNF-2676) identified and categorized a series of potential accidents associated with K Basins Sludge Retrieval System design and operation. The fuel burn accident was of concern with respect to the potential release of contamination resulting from a runaway chemical reaction of the uranium fuel in a knockout pot basket suspended in the air. The unmitigated radiological dose to an offsite receptor from this fuel burn accident is calculated to be much less than the offsite risk evaluation guidelines for anticipated events. However, because of potential radiation exposure to the facility worker, this accident is precluded with a safety significant lifting device that will prevent the monorail hoist from lifting the knockout pot basket out of the K Basin water pool

Classification Of Road Accidents From The Perspective Of Vehicle Safety Systems

Directory of Open Access Journals (Sweden)

Jirovský Václav

2015-11-01

Full Text Available Modern road accident investigation and database structures are focused on accident analysis and classification from the point of view of the accident itself. The presented article offers a new approach, which will describe the accident from the point of view of integrated safety vehicle systems. Seven main categories have been defined to specify the level of importance of automated system intervention. One of the proposed categories is a new approach to defining the collision probability of an ego-vehicle with another object. This approach focuses on determining a 2-D reaction space, which describes all possible positions of the vehicle or other moving object in the specified amount of time in the future. This is to be used for defining the probability of the vehicles interacting - when the intersection of two reaction spaces exists, an action has to be taken on the side of ego-vehicle. The currently used 1-D quantity of TTC (time-to-collision can be superseded by the new reaction space variable. Such new quantity, whose basic idea is described in the article, enables the option of counting not only with necessary braking time, but mitigation by changing direction is then easily feasible. Finally, transparent classification measures of a probable accident are proposed. Their application is highly effective not only during basic accident comparison, but also for an on-board safety system.

Analysis of Three Mile Island - Unit 2 accident

International Nuclear Information System (INIS)

1979-07-01

The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions during Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

Analysis of Three Mile Island-Unit 2 accident

International Nuclear Information System (INIS)

1979-07-01

The Nuclear Safety Analysis Center (NSAC) of the Electic Power Research Institute is analyzing the Three Mile Island-2 accident. An early result of this analysis was a brief narrative summary, issued in mid-May 1979. The present report contains a revised version of that narrative summary, a highly detailed sequence of events, a standard reference list, a list of abbreviations and acronyms, and several appendices. The appendices serve either to describe plant features which are pertinent to the understanding of the sequence of events, or indicate how certain inferences and conclusions in the report were reached. Supplementing the appendices contained herein, additional appendices are in preparation; these will be issued when available (e.g., the appendices Hydrogen Phenomena and Operator Actions duing Initial Transient will follow later). Also in preparation is a matrix of equipment and systems actions during the accident. This report together with future supplements and a separate Core Damage Assessment report, will embody the principal results of that phase of NSAC's work which is devoted to learning and understanding what happened during the accident. Subsequent phases will concentrate on causes, lessons learned and generic remedial or preventive measures which may be appropriate

A CANDU Severe Accident Analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Catana, Alexandru; Prisecaru, Ilie

2006-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents for CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D2O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for 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 temperatures inside a pressure tube reaches 10000 deg C, a contact between pressure tube and calandria tube occurs and the residual 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) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of 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 normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the existing data. The results are encouraging. (authors)

Building of communication system for nuclear accident emergency disposal based on IP multimedia subsystem

Science.gov (United States)

Wang, Kang; Gao, Guiqing; Qin, Yuanli; He, Xiangyong

2018-05-01

The nuclear accident emergency disposal must be supported by an efficient, real-time modularization and standardization communication system. Based on the analysis of communication system for nuclear accident emergency disposal which included many functions such as the internal and external communication, multiply access supporting and command center. Some difficult problems of the communication system were discussed such as variety access device type, complex composition, high mobility, set up quickly, multiply business support, and so on. Taking full advantages of the IP Multimedia Subsystem (IMS), a nuclear accident emergency communication system was build based on the IMS. It was studied and implemented that some key unit and module functions of communication system were included the system framework implementation, satellite access, short-wave access, load/vehicle-mounted communication units. The application tests showed that the system could provide effective communication support for the nuclear accident emergency disposal, which was of great practical value.

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

Energy Technology Data Exchange (ETDEWEB)

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

1997-07-01

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

Preliminary Analysis of a Steam Line Break Accident with the MARS-KS code for the SMART Design with Passive Safety Systems

Energy Technology Data Exchange (ETDEWEB)

Kang, Doohyuk; Ko, Yungjoo; Suh, Jaeseung [Hannam Univ., Daejeon (Korea, Republic of); Bae, Hwang; Ryu, Sunguk; Yi, Sungjae; Park, Hyunsik [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

SMART has been developed by KAERI, and SMART-Standard Design Approval (SDA) was recently granted in 2012. A SMART design with Passive Safety System (PSS) features (called SMART-PSS) is being developed and added to the standard design of SMART by KAERI to improve its safety system. Active safety systems such as safety injection pumps will be replaced by a passive safety system, which is actuated only by the gravity force caused by the height difference. All tanks for the passive safety systems are higher than the injection nozzle, which is located around the reactor coolant pumps (RCPs). In this study, a preliminary analysis of the main steam line break accident (MSLB) was performed using the MARS-KS code to understand the general behavior of the SMART-PSS design and to prepare its validation test with the SMART-ITL (FESTA) facility. An anticipated accident for the main steam line break (MSLB) was performed using the MARS-KS code to understand the thermal-hydraulic behaviors of the SMART-PSS design. The preliminary analysis provides good insight into the passive safety system design features of the SMART-PSS and the thermal-hydraulic characteristics of the SMART design. The analysis results of the MSLB showed that the core water collapsed level inside the core support barrel was maintained high over the active core top level during the transient period. Therefore, the SMART-PSS design has satisfied the requirements to maintain the plant at a safe shutdown condition during 72 hours without AC power or operator action after an anticipated accident.

Accident consequence calculations for project W-058 safety analysis

International Nuclear Information System (INIS)

Van Keuren, J.C.

1997-01-01

Accident consequence analyses have been performed for Project W-058, the Replacement Cross Site Transfer System. using the assumption and analysis techniques developed for the Tank Remediation Waste system Basis for Interim Operation. most potential accident involving the FISTS are bounded by the TWRS BIO analysis. However, the spray leak and pool leak scenarios require revised analyses since the RCSTS design utilizes larger diameter pipe and higher pressures than those analyzed in the TWRS BIO. Also the volume of diversion box and vent station are larger than that assumed for the valve pits in the TWRS BIO, which effects results of sprays or spills into the pits. the revised analysis for the spray leak is presented in Section 2, for the above ground spill in Section 3, for the presented in Section 2, for the above ground spill in Section 3, for the subsurface spill forming a pool in Section 4, and for the subsurface pool remaining subsurface in Section 5. The conclusion from these sections are summarized below

Risk analysis of releases from accidents during mid-loop operation at Surry

International Nuclear Information System (INIS)

Jo, J.; Lin, C.C.; Nimnual, S.; Mubayi, V.; Neymotin, L.

1992-11-01

Studies and operating experience suggest that the risk of severe accidents during low power operation and/or shutdown (LP/S) conditions could be a significant fraction of the risk at full power operation. Two studies have begun at the Nuclear Regulatory Commission (NRC) to evaluate the severe accident progression from a risk perspective during these conditions: One at the Brookhaven National Laboratory for the Surry plant, a pressurized water reactor (PWR), and the other at the Sandia National Laboratories for the Grand Gulf plant, a boiling water reactor (BWR). Each of the studies consists of three linked, but distinct, components: a Level I probabilistic risk analysis (PRA) of the initiating events, systems analysis, and accident sequences leading to core damage; a Level 2/3 analysis of accident progression, fuel damage, releases, containment performance, source term and consequences-off-site and on-site; and a detailed Human Reliability Analysis (HRA) of actions relevant to plant conditions during LP/S operations. This paper summarizes the approach taken for the Level 2/3 analysis at Surry and provides preliminary results on the risk of releases and consequences for one plant operating state, mid-loop operation, during shutdown

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-12-23

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

Accident analysis and DOE criteria

International Nuclear Information System (INIS)

Graf, J.M.; Elder, J.C.

1982-01-01

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

Fatal accidents analysis in Peruvian mining industry

International Nuclear Information System (INIS)

Candia, R. C.; Hennies, W. T.; Azevedo, R. c.; Almeida, I.G.; Soto, J. F.

2010-01-01

Although reductions in the tax of injuries and accidents have been observed in recent years, Mining is still one of the highest risks industries. The basic causes for occurrence of fatalities can be attributed to unsafe conditions and unsafe acts. In this scene is necessary to identify safety problems and to aim the effective solutions. On the other hand, the developing countries dependence on primary industries as mining is evident. In the Peruvian economy, approximately 16% of the GNP and more than 50% of the exportations are due to the mining sector, detaching its competitive position in the worldwide mining. This paper presents fatal accidents analysis in the Peruvian mining industry, having as basis the register of occurred fatal accidents since year 2000 until 2007, identifying the main types of accidents occurred. The source of primary information is the General Mining Direction (DGM) of the Peruvian Mining and Energy Ministry (MEM). The majority of victims belongs to tertiary contractor companies that render services for mine companies. The results of the analysis show also that the majority of accidents happened in the underground mines, and that it is necessary to propose effective solutions to manage risks, aiming at reducing the fatal accidents taxes. (Author)

Analysis of small break loss of coolant accident for Chinese CPR1000

Energy Technology Data Exchange (ETDEWEB)

Kim, Ju Youl [FNC Technology Co., Yongin (Korea, Republic of); Cilier, Anthonie [North-West University, Mahikeng (South Africa); Poc, Li-chi Cliff [Micro-Simulation Technology, Montville (United States)

2016-05-15

This research analyses the small break loss of coolant accident (LOCA) on a Chinese CPR1000 type reactor. LOCA accident is used as benchmark for the PCTRAN/CPR1000 code by comparing the effects and results to the Manshaan FSAR accident analysis. LOCA is a design basis accident in which a guillotine break is postulated to occur in one of the cold legs of a pressurized water reactor (PWR). Consequently, the primary system pressure would drop and almost all the reactor coolant would be discharged into the reactor containment. The drop in pressure would activate the reactor protection system and the reactor would trip. The simulation of a 3-inch small break loss of coolant accident using the PCTRAN/CPR1000 has revealed this code's effectiveness as well as weaknesses in specific simulation applications. The code has the ability to run at 16 times real time and produce very accurate results. The results are consistently producing the same trends as licensed codes used in Safety Assessment Reports. It is however able to produce these results in a fraction of the time and also provides a whole plant simulation coupling the various thermal, hydraulic, chemical and neutronic systems together with a plant specific control system.

An analysis of the Three Mile Island accident

International Nuclear Information System (INIS)

Brooks, G.L.; Siddal, E.

1980-09-01

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

Human reliability data, human error and accident models--illustration through the Three Mile Island accident analysis

International Nuclear Information System (INIS)

Le Bot, Pierre

2004-01-01

Our first objective is to provide a panorama of Human Reliability data used in EDF's Safety Probabilistic Studies, and then, since these concepts are at the heart of Human Reliability and its methods, to go over the notion of human error and the understanding of accidents. We are not sure today that it is actually possible to provide in this field a foolproof and productive theoretical framework. Consequently, the aim of this article is to suggest potential paths of action and to provide information on EDF's progress along those paths which enables us to produce the most potentially useful Human Reliability analyses while taking into account current knowledge in Human Sciences. The second part of this article illustrates our point of view as EDF researchers through the analysis of the most famous civil nuclear accident, the Three Mile Island unit accident in 1979. Analysis of this accident allowed us to validate our positions regarding the need to move, in the case of an accident, from the concept of human error to that of systemic failure in the operation of systems such as a nuclear power plant. These concepts rely heavily on the notion of distributed cognition and we will explain how we applied it. These concepts were implemented in the MERMOS Human Reliability Probabilistic Assessment methods used in the latest EDF Probabilistic Human Reliability Assessment. Besides the fact that it is not very productive to focus exclusively on individual psychological error, the design of the MERMOS method and its implementation have confirmed two things: the significance of qualitative data collection for Human Reliability, and the central role held by Human Reliability experts in building knowledge about emergency operation, which in effect consists of Human Reliability data collection. The latest conclusion derived from the implementation of MERMOS is that, considering the difficulty in building 'generic' Human Reliability data in the field we are involved in, the best

Development of a totally integrated severe accident training system

International Nuclear Information System (INIS)

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

2006-01-01

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

Analysis of labor accidents in Brazil, 2004-2007

OpenAIRE

Alves, Everton Fernando

2010-01-01

Current research synthesizes epidemiological data on morbo- mortality by labor accidents in the Brazilian population and gives a cross- section of these accidents in Brazil between 2004 and 2007. Current descrip- tive and exploratory analysis uses databases of thePublic Health Ministry on labor accidents. In fact, 465.700 and 653.090 labor accidents were notified respectively in 2004 and 2007, with a trend towardsan increase in number. The state of Santa Catarina was t...

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

International Nuclear Information System (INIS)

Hill, Axel

2014-01-01

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

Recursive modeling of loss of control in human and organizational processes: a systemic model for accident analysis.

Science.gov (United States)

Kontogiannis, Tom; Malakis, Stathis

2012-09-01

A recursive model of accident investigation is proposed by exploiting earlier work in systems thinking. Safety analysts can understand better the underlying causes of decision or action flaws by probing into the patterns of breakdown in the organization of safety. For this deeper analysis, a cybernetic model of organizational factors and a control model of human processes have been integrated in this article (i.e., the viable system model and the extended control model). The joint VSM-ECOM framework has been applied to a case study to help safety practitioners with the analysis of patterns of breakdown with regard to how operators and organizations manage goal conflicts, monitor work progress, recognize weak signals, align goals across teams, and adapt plans on the fly. The recursive accident representation brings together several organizational issues (e.g., the dilemma of autonomy versus compliance, or the interaction between structure and strategy) and addresses how operators adapt to challenges in their environment by adjusting their modes of functioning and recovery. Finally, it facilitates the transfer of knowledge from diverse incidents and near misses within similar domains of practice. Copyright © 2012 Elsevier Ltd. All rights reserved.

The covariance between the number of accidents and the number of victims in multivariate analysis of accident related outcomes

NARCIS (Netherlands)

Bijleveld, F. D.

In this study some statistical issues involved in the simultaneous analysis of accident related outcomes of the road traffic process are investigated. Since accident related outcomes like the number of victims, fatalities or accidents show interdependencies, their simultaneous analysis requires that

[A spatially explicit analysis of traffic accidents involving pedestrians and cyclists in Berlin].

Science.gov (United States)

Lakes, Tobia

2017-12-01

In many German cities and counties, sustainable mobility concepts that strengthen pedestrian and cyclist traffic are promoted. From the perspectives of urban development, traffic planning and public healthcare, a spatially differentiated analysis of traffic accident data is decisive. 1) The identification of spatial and temporal patterns of the distribution of accidents involving cyclists and pedestrians, 2) the identification of hotspots and exploration of possible underlying causes and 3) the critical discussion of benefits and challenges of the results and the derivation of conclusions. Spatio-temporal distributions of data from accident statistics in Berlin involving pedestrians and cyclists from 2011 to 2015 were analysed with geographic information systems (GIS). While the total number of accidents remains relatively stable for pedestrian and cyclist accidents, the spatial distribution analysis shows, however, that there are significant spatial clusters (hotspots) of traffic accidents with a strong concentration in the inner city area. In a critical discussion, the benefits of geographic concepts are identified, such as spatially explicit health data (in this case traffic accident data), the importance of the integration of other data sources for the evaluation of the health impact of areas (traffic accident statistics of the police), and the possibilities and limitations of spatial-temporal data analysis (spatial point-density analyses) for the derivation of decision-supported recommendations and for the evaluation of policy measures of health prevention and of health-relevant urban development.

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

International Nuclear Information System (INIS)

Wider, H.; Cametti, J.; Clusaz, A.; Devos, J.; VanGoethem, G.; Nguyen, H.; Sola, A.

1985-01-01

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

Fast reactor fuel failures and steam generator leaks: Transient and accident analysis approaches

International Nuclear Information System (INIS)

1996-10-01

This report consists of a survey of activities on transient and accident analysis for the LMFR. It is focused on the following subjects: Fuel transient tests and analyses in hypothetical incident/accident situations; sodium-water interaction in steam generators, and sodium fires: test and analyses. There are also sections dealing with the experimental and analytical studies of: fuel subassembly failures; sodium boiling, molten fuel-coolant interaction; molten material movement and relocation in fuel bundles; heat removal after an accident or incident; sodium-water reaction in steam generator; steam generator protection systems; sodium-water contact in steam generator building; fire-fighting methods and systems to deal with sodium fires. Refs, figs, tabs

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

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

2000-01-01

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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

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

Factors correlated with traffic accidents as a basis for evaluating Advanced Driver Assistance Systems.

Science.gov (United States)

Staubach, Maria

2009-09-01

This study aims to identify factors which influence and cause errors in traffic accidents and to use these as a basis for information to guide the application and design of driver assistance systems. A total of 474 accidents were examined in depth for this study by means of a psychological survey, data from accident reports, and technical reconstruction information. An error analysis was subsequently carried out, taking into account the driver, environment, and vehicle sub-systems. Results showed that all accidents were influenced by errors as a consequence of distraction and reduced activity. For crossroad accidents, there were further errors resulting from sight obstruction, masked stimuli, focus errors, and law infringements. Lane departure crashes were additionally caused by errors as a result of masked stimuli, law infringements, expectation errors as well as objective and action slips, while same direction accidents occurred additionally because of focus errors, expectation errors, and objective and action slips. Most accidents were influenced by multiple factors. There is a safety potential for Advanced Driver Assistance Systems (ADAS), which support the driver in information assimilation and help to avoid distraction and reduced activity. The design of the ADAS is dependent on the specific influencing factors of the accident type.

Accident analysis in nuclear power plants

International Nuclear Information System (INIS)

Silva, D.E. da

1981-01-01

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

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

The Specialists' Meeting on Gas-Cooled Reactor Safety and Accident Analysis was convened by the International Atomic Energy Agency in Oak Ridge on the invitation of the Department of Energy in Washington, USA. The meeting was hosted by the Oak Ridge National Laboratory. The purpose of the meeting was to provide an opportunity to compare and discuss results of safety and accident analysis of gas-cooled reactors under development, construction or in operation, to review their lay-out, design, and their operational performance, and to identify areas in which additional research and development are needed. The meeting emphasized the high safety margins of gas-cooled reactors and gave particular attention to the inherent safety features of small reactor units. The meeting was subdivided into four technical sessions: Safety and Related Experience with Operating Gas-Cooled Reactors (4 papers); Risk and Safety Analysis (11 papers); Accident Analysis (9 papers); Miscellaneous Related Topics (5 papers). A separate abstract was prepared for each of these papers

PWR systems transient analysis

International Nuclear Information System (INIS)

Kennedy, M.F.; Peeler, G.B.; Abramson, P.B.

1985-01-01

Analysis of transients in pressurized water reactor (PWR) systems involves the assessment of the response of the total plant, including primary and secondary coolant systems, steam piping and turbine (possibly including the complete feedwater train), and various control and safety systems. Transient analysis is performed as part of the plant safety analysis to insure the adequacy of the reactor design and operating procedures and to verify the applicable plant emergency guidelines. Event sequences which must be examined are developed by considering possible failures or maloperations of plant components. These vary in severity (and calculational difficulty) from a series of normal operational transients, such as minor load changes, reactor trips, valve and pump malfunctions, up to the double-ended guillotine rupture of a primary reactor coolant system pipe known as a Large Break Loss of Coolant Accident (LBLOCA). The focus of this paper is the analysis of all those transients and accidents except loss of coolant accidents

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

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

Passive Decay Heat Removal Strategy of Integrated Passive Safety System (IPSS) for SBO-combined Accidents

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Ho; Chang, Soon Heung; Jeong, Yong Hoon [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

The weak points of nuclear safety would be in outmoded nuclear power plants like the Fukushima reactors. One of the systems for the safety enhancement is integrated passive safety system (IPSS) proposed after the Fukushima accidents. It has the five functions for the prevention and mitigation of a severe accident. Passive decay heat removal (PDHR) strategy using IPSS is proposed for coping with SBO-combined accidents in this paper. The two systems for removing decay heat before core-melt were applied in the strategy. The accidents were simulated by MARS code. The reference reactor was OPR1000, specifically Ulchin-3 and 4. The accidents included loss-of-coolant accidents (LOCA) because the coolant losses could be occurred in the SBO condition. The examples were the stuck open of PSV, the abnormal open of SDV and the leakage of RCP seal water. Also, as LOCAs with the failure of active safety injection systems were considered, various LOCAs were simulated in SBO. Based on the thermal hydraulic analysis, the probabilistic safety analysis was carried out for the PDHR strategy to estimate the safety enhancement in terms of the variation of core damage frequency. AIMS-PSA developed by KAERI was used for calculating CDF of the plant. The IPSS was applied in the PDHR strategy which was developed in order to cope with the SBO-combined accidents. The estimation for initiating SGGI or PSIS was based on the pressure in RCS. The simulations for accidents showed that the decay heat could be removed for the safety duration time in SBO. The increase of safety duration time from the strategy provides the increase of time for the restoration of AC power.

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

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jong Tae; Jung, Won Dea

2001-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Jeong, Jong Tae; Jung, Won Dea [Korea Atomic Energy Research Institute, Taejeon (Korea)

2002-03-01

The computerized system, NARIS(Nuclear Accident Risk Information System) was developed in order to support the estimation of health effects and the establishment the effective risk reduction strategies. Using the system, we can analyze the distribution of health effects easily by displaying the results on the digital map of the site. Also, the thematic mapping allows the diverse analysis of the distribution of the health effects.The NARIS can be used in the emergency operation facilities in order to analyze the distribution of the health effects resulting from the severe accidents of a nuclear power plant. Also, the rapid analysis of the health effect is possible by storing the health effect results in the form of a database. Therefore, the staffs of the emergency operation facilities can establish the rapid and effective emergency response strategies. The module for the optimization of the costs and benefits and the decision making support will be added. The technical support for the establishment of the optimum and effective emergency response strategies will be possible using this system. 23 figs., 1 tab. (Author)

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

International Nuclear Information System (INIS)

Jeong, Jong Tae; Jung, Won Dea

2001-03-01

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

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

Energy Technology Data Exchange (ETDEWEB)

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

2017-06-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Application of the Severe Accident Code ATHLET-CD. Coolant injection to primary circuit of a PWR by mobile pump system in case of SBLOCA severe accident scenario

International Nuclear Information System (INIS)

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

2017-01-01

The improvement of the safety of nuclear power plants is a continuously on-going process. The analysis of transients and accidents is an important research topic, which significantly contributes to safety enhancements of existing power plants. In case of an accident with multiple failures of safety systems, core uncovery and heat-up can occur. In order to prevent the accident to turn into a severe one or to mitigate the consequences of severe accidents, different accident management measures can be applied. By means of numerical analyses performed with the compute code ATHLET-CD, the effectiveness of coolant injection with a mobile pump system into the primary circuit of a PWR was studied. According to the analyses, such a system can stop the melt progression if it is activated prior to 10 % of total core is molten.

Energy Analysis of Road Accidents Based on Close-Range Photogrammetry

Directory of Open Access Journals (Sweden)

Alejandro Morales

2015-11-01

Full Text Available This paper presents an efficient and low-cost approach for energy analysis of road accidents using images obtained using consumer-grade digital cameras and smartphones. The developed method could be used by security forces in order to improve the qualitative and quantitative analysis of traffic accidents. This role of the security forces is crucial to settle arguments; consequently, the remote and non-invasive collection of accident related data before the scene is modified proves to be essential. These data, taken in situ, are the basis to perform the necessary calculations, basically the energy analysis of the road accident, for the corresponding expert reports and the reconstruction of the accident itself, especially in those accidents with important damages and consequences. Therefore, the method presented in this paper provides the security forces with an accurate, three-dimensional, and scaled reconstruction of a road accident, so that it may be considered as a support tool for the energy analysis. This method has been validated and tested with a real crash scene simulated by the local police in the Academy of Public Safety of Extremadura, Spain.

Radionuclides release possibility analysis of MSR at various accident conditions

Energy Technology Data Exchange (ETDEWEB)

Lee, Choong Wie; Kim, Hee Reyoung [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

There are some accidents which go beyond our expectation such as Fukushima Daiichi nuclear disaster and amounts of radionuclides release to environment, so more effort and research are conducted to prevent it. MSR (Molten Salt Reactor) is one of GEN-IV reactor types, and its coolant and fuel are mixtures of molten salt. MSR has a schematic like figure 1 and it has different features with the solid fuel reactor, but most important and interesting feature of MSR is its many safety systems. For example, MSR has a large negative void coefficient. Even though power increases, the reactor slows down soon. Radionuclides release possibility of MSR was analyzed at various accident conditions including Chernobyl and Fukushima ones. The MSR was understood to prevent the severe accident by the negative reactivity coefficient and the absence of explosive material such as water at the Chernobyl disaster condition. It was expected to contain fuel salts in the reactor building and not to release radionuclides into environment even if the primary system could be ruptured or broken and fuel salts would be leaked at the Fukushima Daiichi nuclear disaster condition of earthquake and tsunami. The MSR, which would not lead to the severe accident and therefore prevents the fuel release to the environment at many expected scenarios, was thought to have priority in the aspect of accidents. A quantitative analysis and a further research are needed to evaluate the possibility of radionuclide release to the environment at the various accident conditions based on the simple comparison of the safety feature between MSR and solid fuel reactor.

[Model of Analysis and Prevention of Accidents - MAPA: tool for operational health surveillance].

Science.gov (United States)

de Almeida, Ildeberto Muniz; Vilela, Rodolfo Andrade de Gouveia; da Silva, Alessandro José Nunes; Beltran, Sandra Lorena

2014-12-01

The analysis of work-related accidents is important for accident surveillance and prevention. Current methods of analysis seek to overcome reductionist views that see these occurrences as simple events explained by operator error. The objective of this paper is to analyze the Model of Analysis and Prevention of Accidents (MAPA) and its use in monitoring interventions, duly highlighting aspects experienced in the use of the tool. The descriptive analytical method was used, introducing the steps of the model. To illustrate contributions and or difficulties, cases where the tool was used in the context of service were selected. MAPA integrates theoretical approaches that have already been tried in studies of accidents by providing useful conceptual support from the data collection stage until conclusion and intervention stages. Besides revealing weaknesses of the traditional approach, it helps identify organizational determinants, such as management failings, system design and safety management involved in the accident. The main challenges lie in the grasp of concepts by users, in exploring organizational aspects upstream in the chain of decisions or at higher levels of the hierarchy, as well as the intervention to change the determinants of these events.

Severe accident analysis and management in nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Golshan, Mina

2013-01-01

Within the UK regulatory regime, assessment of risks arising from licensee's activities are expected to cover both normal operations and fault conditions. In order to establish the safety case for fault conditions, fault analysis is expected to cover three forms of analysis: design basis analysis (DBA), probabilistic safety assessment (PSA) and severe accident analysis (SAA). DBA should provide a robust demonstration of the fault tolerance of the engineering design and the effectiveness of the safety measures on a conservative basis. PSA looks at a wider range of fault sequences (on a best estimate basis) including those excluded from the DBA. SAA considers significant but unlikely accidents and provides information on their progression and consequences, within the facility, on the site and off site. The assessment of severe accidents is not limited to nuclear power plants and is expected to be carried out for all plant states where the identified dose targets could be exceeded. This paper sets out the UK nuclear regulatory expectation on what constitutes a severe accident, irrespective of the type of facility, and describes characteristics of severe accidents focusing on nuclear fuel cycle facilities. Key rules in assessment of severe accidents as well as the relationship to other fault analysis techniques are discussed. The role of SAA in informing accident management strategies and offsite emergency plans is covered. The paper also presents generic examples of scenarios that could lead to severe accidents in a range of nuclear fuel cycle facilities. (authors)

A cluster analysis on road traffic accidents using genetic algorithms

Science.gov (United States)

Saharan, Sabariah; Baragona, Roberto

2017-04-01

The analysis of traffic road accidents is increasingly important because of the accidents cost and public road safety. The availability or large data sets makes the study of factors that affect the frequency and severity accidents are viable. However, the data are often highly unbalanced and overlapped. We deal with the data set of the road traffic accidents recorded in Christchurch, New Zealand, from 2000-2009 with a total of 26440 accidents. The data is in a binary set and there are 50 factors road traffic accidents with four level of severity. We used genetic algorithm for the analysis because we are in the presence of a large unbalanced data set and standard clustering like k-means algorithm may not be suitable for the task. The genetic algorithm based on clustering for unknown K, (GCUK) has been used to identify the factors associated with accidents of different levels of severity. The results provided us with an interesting insight into the relationship between factors and accidents severity level and suggest that the two main factors that contributes to fatal accidents are "Speed greater than 60 km h" and "Did not see other people until it was too late". A comparison with the k-means algorithm and the independent component analysis is performed to validate the results.

Development of an accident management expert system for containment assessment

International Nuclear Information System (INIS)

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

1987-01-01

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

Development of Draft Regulatory Guide on Accident Analysis for Nuclear Power Plants with New Safety Design Features

Energy Technology Data Exchange (ETDEWEB)

Bang, Young Seok; Woo, Sweng Woong; Hwang, Tae Suk [KINS, Daejeon (Korea, Republic of); Sim, Suk K; Hwang, Min Jeong [Environment and Energy Technology, Daejeon (Korea, Republic of)

2016-05-15

The present paper discusses the development process of the draft version of regulatory guide (DRG) on accident analysis of the NPP having the NSFD and its result. Based on the consideration on the lesson learned from the previous licensing review, a draft regulatory guide (DRG) on accident analysis for NPP with new safety design features (NSDF) was developed. New safety design features (NSDF) have been introduced to the new constructing nuclear power plants (NPP) since the early 2000 and the issuance of construction permit of SKN Units 3 and 4. Typical examples of the new safety features includes Fluidic Device (FD) within Safety Injection Tanks (SIT), Passive Auxiliary Feedwater System (PAFS), ECCS Core Barrel Duct (ECBD) which were adopted in APR1400 design and/or APR+ design to improve the safety margin of the plants for the postulated accidents of interest. Also several studies of new concept of the safety system such as Hybrid ECCS design have been reported. General and/or specific guideline of accident analysis considering the NSDF has been requested. Realistic evaluation of the impact of NSDF on accident with uncertainty and separated accident analysis accounting the NSDF impact were specified in the DRG. Per the developmental process, identification of key issues, demonstration of the DRG with specific accident with specific NSDF, and improvement of DGR for the key issues and their resolution will be conducted.

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

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

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

International Nuclear Information System (INIS)

Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

2009-07-01

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

Analysis of unmitigated large break loss of coolant accidents using MELCOR code

Science.gov (United States)

Pescarini, M.; Mascari, F.; Mostacci, D.; De Rosa, F.; Lombardo, C.; Giannetti, F.

2017-11-01

In the framework of severe accident research activity developed by ENEA, a MELCOR nodalization of a generic Pressurized Water Reactor of 900 MWe has been developed. The aim of this paper is to present the analysis of MELCOR code calculations concerning two independent unmitigated large break loss of coolant accident transients, occurring in the cited type of reactor. In particular, the analysis and comparison between the transients initiated by an unmitigated double-ended cold leg rupture and an unmitigated double-ended hot leg rupture in the loop 1 of the primary cooling system is presented herein. This activity has been performed focusing specifically on the in-vessel phenomenology that characterizes this kind of accidents. The analysis of the thermal-hydraulic transient phenomena and the core degradation phenomena is therefore here presented. The analysis of the calculated data shows the capability of the code to reproduce the phenomena typical of these transients and permits their phenomenological study. A first sequence of main events is here presented and shows that the cold leg break transient results faster than the hot leg break transient because of the position of the break. Further analyses are in progress to quantitatively assess the results of the code nodalization for accident management strategy definition and fission product source term evaluation.

Indonesian Sea Accident Analysis (Case Study From 2003 – 2013)

Science.gov (United States)

Arya Dewanto, Y.; Faturachman, D.

2018-03-01

There are so many accidents in sea transportation in Indonesia. Most of the accidents happen because of low concern aspects of the safety and security of the crew. In sailing, a man as transport users to interact with the ship and the surrounding environment (including other ships, cruise lines, ports, and the situation of local conditions). These interactions are sometimes very complex and related to various aspects of. Aware of the multiplicity of aspects related to the third of these factors, seeking the safety of cruise through a reduction in the number of accidents and the risk of death and serious injuries due to accidents and goods transported is certainly not enough attempted through mono-sector approach, but rather takes a multi-sector approach to the efforts. In this paper, we described the Indonesian Sea Transportation accident analysis for eleven years divided into four items: total of ship accident type, ship accident factor, total of casualties, region of ship accidents. All data founded from Marine Court (Mahkamah Pelayaran). From that 4 items we can find Indonesia Sea Accident Analysis from 2003-2013.

Study on the experimental VHTR safety with analysis for a hypothetical rapid depressurization accident

International Nuclear Information System (INIS)

Mitake, S.; Suzuki, K.; Ohno, T.; Okada, T.

1982-01-01

A hypothetical rapid depressurization accident of the experimental VHTR has been analyzed, including all phenomena in the accident, from its initiating depressurization of the coolant to consequential radiological hazard. Based on reliability analysis of the engineered safety features, all possible sequences, in which the safety systems are in success or in failure, have been investigated with event tree analysis. The result shows the inherent safety characteristics of the reactor and the effectiveness of the engineered safety features. And through the analysis, it has been indicated that further investigations on some phenomena in the accident, e.g., air ingress by natural circulation flow and fission product transport in the plant, will bring forth more reasonable and sufficient safety of the reactor

International cooperation in accident analysis of RBMK reactors

International Nuclear Information System (INIS)

Kaliatka, A.; Isag

2005-01-01

Chouha Michel (Institute for Radiological Protection and Nuclear Safety), D'Auria Francesco (Institute of Pisa), Kaliatka Algirdas (Lithuanian Energy Institute), Uspuras Eugenijus (Lithuanian Energy Institute). The safety of nuclear power plants is a primary concern of the European Union (EU) and its Member States. In the early 1990s, the European Union decided to take a prominent role in international efforts to help the New Independent States (NIS) and countries of central Europe to ensure the safety of their nuclear reactors. The Commission's approach to nuclear safety in central and Eastern Europe and the NIS is based on two main objectives, which are fully in line with the policy of the international community as decided by the G7 in 1992: 1) In the short term, to improve operational safety; to make near term technical improvements to plants based on safety assessments and to enhance regulatory regimes; 2) In the longer term, to examine the scope for replacing less safe plants by the development of alternative energy sources and more efficient use of energy and to examine the potential for upgrading plants of more recent design. In this paper the safety concerns, related to RBMK type reactors (Russian acronym for 'Channelized Large Power Reactor) are discussed. These reactors were not exported and were built exclusively in the territory of the former Soviet Union. There are presently plants at Saint Petersburg (Sosnovy Bor), Kursk, Chernobyl and Smolensk. A total of 17 such reactors have been built and 12 are currently in operation. Two international projects: TACIS project 'Development of a code system for severe accident analysis in RBMK reactors' and PHARE projects 'Support to VATESI for Important Tasks Relevant to the Licensing Activities of Ignalina Nuclear Power Plant' are presented. The aim of the TACIS project is to help the Russian Authorities to build such capabilities, for their RBMK nuclear power plants (NPPs). The drawing of the Tacis nuclear

Preliminary evaluation of the Accident Response Mobile Manipulation System for accident site salvage operations

International Nuclear Information System (INIS)

Trujillo, J.M.; Morse, W.D.; Jones, D.P.

1994-01-01

This paper describes and evaluates operational experiences with the Accident Response Mobile Manipulation System (ARMMS) during simulated accident site salvage operations which might involve nuclear weapons. The ARMMS is based upon a teleoperated mobility platform with two Schilling Titan 7F Manipulators

Macro Data Analysis of Traffic Accidents in Indonesia

Directory of Open Access Journals (Sweden)

Annisa Jusuf

2017-04-01

Full Text Available This paper presents a macro data analysis of Indonesian road accidents in the form of statistical data. Traffic accidents and their subsequent fatalities bring enormous social and economic consequences. A good understanding of the problem is expected to initiate major action toward the improvement of road and vehicle safety. One important milestone is the collection and analysis of road accident data. The results from this study portray the ‘tangled threads’ problem of traffic in Indonesia. The population number and number of vehicles have increased steadily, as has been accurately predicted by experts. Meanwhile, there is not enough infrastructure growth. Motorcycles are the main contributor to traffic accidents and fatalities due to their popularity as an effective vehicle to jump traffic jams. The ‘tangled threads’ need an extremely creative and comprehensive solution.

Improvements in the nuclear accident response system in Brazil

International Nuclear Information System (INIS)

Estrada, J.J.S.; Azevedo, E.M.; Knofel, T.M.J.; Recio, J.C.A.; Alves, R.N.

1998-01-01

The National Commission on Nuclear Energy has been making outstanding effort to improve its nuclear and radiological accident response systems since the tragic accident in Goiania. Most of this effort is related to nuclear area although the radiological accident has been also considered. This paper describes the improvements in the CNEN response system structure, discusses several topics involving those related to emergency planning and preparedness, and points out some deficiencies that need to be corrected also. The situation during the Goiania accident was more disadvantageous than nowadays, so it is believed that none of the actual deficiencies are sufficient to guess that the population and the environment will not be protected in case of a nuclear or radiological accident

High enrichment to low enrichment core's conversion. Accidents analysis

International Nuclear Information System (INIS)

Abbate, P.; Rubio, R.; Doval, A.; Lovotti, O.

1990-01-01

This work analyzes the different accidents that may occur in the reactor's facility after the 20% high-enriched uranium core's conversion. The reactor (of 5 thermal Mw), built in the 50's and 60's, is of the 'swimming pool' type, with light water and fuel elements of the curve plates MTR type, enriched at 93.15 %. This analysis includes: a) accidents by reactivity insertion; b) accidents by coolant loss; c) analysis by flow loss and d) fission products release. (Author) [es

System calculations related to the accident at Three-Mile Island using TRAC

International Nuclear Information System (INIS)

Ireland, J.R.

1980-01-01

The Three Mile Island nuclear plant (Unit 2) was modeled using the Transient Reactor Analysis Code (TRAC-P1A) and a base case calculation, which simulated the initial part of the accident that occurred on March 28, 1979, was performed. In addition to the base case calculation, several parametric calculations were performed in which a single hypothetical change was made in the system conditions, such as assuming the high pressure injection (HPI) system operated as designed rather than as in the accident. Some of the important system parameter comparisons for the base case as well as some of the parametric case results are presented

Analysis of reactivity accidents in PWR'S

International Nuclear Information System (INIS)

Camous, F.; Chesnel, A.

1989-12-01

This note describes the French strategy which has consisted, firstly, in examining all the accidents presented in the PWR unit safety reports in order to determine for each parameter the impact on accident consequences of varying the parameter considered, secondly in analyzing the provisions taken into account to restrict variation of this parameter to within an acceptable range and thirdly, in checking that the reliability of these provisions is compatible with the potential consequences of transgression of the authorized limits. Taking into consideration violations of technical operating specifications and/or non-observance of operating procedures, equipment failures, and partial or total unavailability of safety systems, these studies have shown that fuel mechanical strength limits can be reached but that the probability of occurrence of the corresponding events places them in the residual risk field and that it must, in fact, be remembered that there is a wide margin between the design basis accidents and accidents resulting in fuel destruction. However, during the coming year, we still have to analyze scenarios dealing with cumulated events or incidents leading to a reactivity accident. This program will be mainly concerned with the impact of the cases examined relating to dilution incidents under normal operating conditions or accident operating conditions

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

International Nuclear Information System (INIS)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J.

2001-01-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

Application of the coupled Relap5/Panther codes for PWR steam. Line break accident analysis

Energy Technology Data Exchange (ETDEWEB)

Guisset, J.-P.; Bosso, S.; Charlier, A.; Delhaye, X.; Ergo, O.; Ouliddren, K.; Schneidesch, C.; Zhang, J. [Tractebel Energy Engineering, Brussels (Belgium)

2001-07-01

A dynamic coupling between the existing 1-dimensional thermal-hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER is applied via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel thermal-hydraulic analysis code COBRA 3C allows direct evaluation of the Departure from Nucleate Boiling Ratio in parallel with the coupled PANTHER/RELAP5 simulation. The coupled codes are applied to develop a Final Safety Analysis Report (FSAR) accident analysis methodology for the major Steam Line Break (SLB) accident at hot zero power in a typical three-loop pressurised water reactor. In this methodology, the uncertainties related to the plant, core thermal-hydraulic and neutronic parameters are combined in a deterministic bounding approach based on sensitivity studies. The results of coupled thermal-hydraulic and neutronic analysis of SLB are presented and discussed. It is shown that there exists an important margin in the traditional FSAR accident analysis for SLB, which can be attributed by the conservatism's introduced by de-coupling the plant sub-systems. (author)

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

International Nuclear Information System (INIS)

Kaliatka, A.; Valinčius, M.

2012-01-01

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

Analysis of construction accidents in Spain, 2003-2008.

Science.gov (United States)

López Arquillos, Antonio; Rubio Romero, Juan Carlos; Gibb, Alistair

2012-12-01

The research objective for this paper is to obtain a new extended and updated insight to the likely causes of construction accidents in Spain, in order to identify suitable mitigating actions. The paper analyzes all construction sector accidents in Spain between 2003 and 2008. Ten variables were chosen and the influence of each variable is evaluated with respect to the severity of the accident. The descriptive analysis is based on a total of 1,163,178 accidents. Results showed that the severity of accidents was related to variables including age, CNAE (National Classification of Economic Activities) code, size of company, length of service, location of accident, day of the week, days of absence, deviation, injury, and climatic zones. According to data analyzed, a large company is not always necessarily safer than a small company in the aspect of fatal accidents, experienced workers do not have the best accident fatality rates, and accidents occurring away from the usual workplace had more severe consequences. Results obtained in this paper can be used by companies in their occupational safety strategies, and in their safety training programs. Copyright © 2012 National Safety Council and Elsevier Ltd. All rights reserved.

APT Blanket System Loss-of-Flow Accident (LOFA) Analysis Based on Initial Conceptual Design - Case 1: with Beam Shutdown and Active RHR

International Nuclear Information System (INIS)

Hamm, L.L.

1998-01-01

This report is one of a series of reports that document normal operation and accident simulations for the Accelerator Production of Tritium (APT) blanket heat removal system. These simulations were performed for the Preliminary Safety Analysis Report

Modular Accident Analysis Program (MAAP) - MELCOR Crosswalk: Phase II Analyzing a Partially Recovered Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Andrews, Nathan [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Faucett, Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Haskin, Troy Christopher [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Luxat, Dave [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Geiger, Garrett [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Codella, Brittany [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-10-01

Following the conclusion of the first phase of the crosswalk analysis, one of the key unanswered questions was whether or not the deviations found would persist during a partially recovered accident scenario, similar to the one that occurred in TMI - 2. In particular this analysis aims to compare the impact of core degradation morphology on quenching models inherent within the two codes and the coolability of debris during partially recovered accidents. A primary motivation for this study is the development of insights into how uncertainties in core damage progression models impact the ability to assess the potential for recovery of a degraded core. These quench and core recovery models are of the most interest when there is a significant amount of core damage, but intact and degraded fuel still remain in the cor e region or the lower plenum. Accordingly this analysis presents a spectrum of partially recovered accident scenarios by varying both water injection timing and rate to highlight the impact of core degradation phenomena on recovered accident scenarios. This analysis uses the newly released MELCOR 2.2 rev. 966 5 and MAAP5, Version 5.04. These code versions, which incorporate a significant number of modifications that have been driven by analyses and forensic evidence obtained from the Fukushima - Daiichi reactor site.

The computer aided education and training system for accident management

International Nuclear Information System (INIS)

Yoneyama, Mitsuru; Masuda, Takahiro; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

2000-01-01

Under severe accident conditions of a nuclear power plant, plant operators and technical support center (TSC) staffs will be under a amount of stress. Therefore, those individuals responsible for managing the plant should promote their understanding about the accident management and operations. Moreover, it is also important to train in ordinary times, so that they can carry out accident management operations effectively on severe accidents. Therefore, the education and training system which works on personal computers was developed by Japanese BWR group (Tokyo Electric Power Co.,Inc., Tohoku Electric Power Co. ,Inc., Chubu Electric Power Co. ,Inc., Hokuriku Electric Power Co.,Inc., Chugoku Electric Power Co.,Inc., Japan Atomic Power Co.,Inc.), and Hitachi, Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the other is education and training system with a computer simulation. Both systems are designed to execute on MS-Windows(R) platform of personal computers. These systems provide plant operators and technical support center staffs with an effective education and training tool for accident management. TEPCO used the simulation system for the emergency exercise assuming the occurrence of hypothetical severe accident, and have performed an effective exercise in March, 2000. (author)

Visualization of Traffic Accidents

Science.gov (United States)

Wang, Jie; Shen, Yuzhong; Khattak, Asad

2010-01-01

Traffic accidents have tremendous impact on society. Annually approximately 6.4 million vehicle accidents are reported by police in the US and nearly half of them result in catastrophic injuries. Visualizations of traffic accidents using geographic information systems (GIS) greatly facilitate handling and analysis of traffic accidents in many aspects. Environmental Systems Research Institute (ESRI), Inc. is the world leader in GIS research and development. ArcGIS, a software package developed by ESRI, has the capabilities to display events associated with a road network, such as accident locations, and pavement quality. But when event locations related to a road network are processed, the existing algorithm used by ArcGIS does not utilize all the information related to the routes of the road network and produces erroneous visualization results of event locations. This software bug causes serious problems for applications in which accurate location information is critical for emergency responses, such as traffic accidents. This paper aims to address this problem and proposes an improved method that utilizes all relevant information of traffic accidents, namely, route number, direction, and mile post, and extracts correct event locations for accurate traffic accident visualization and analysis. The proposed method generates a new shape file for traffic accidents and displays them on top of the existing road network in ArcGIS. Visualization of traffic accidents along Hampton Roads Bridge Tunnel is included to demonstrate the effectiveness of the proposed method.

[Analysis of accidents for magnetically induced displacement of the large ferromagnetic material in magnetic resonance systems].

Science.gov (United States)

Yamatani, Yuya; Doi, Tsukasa; Ueyama, Tsuyoshi; Nishiki, Shigeo; Ogura, Akio; Kawamitsu, Hideaki; Tsuchihashi, Toshio; Okuaki, Tomoyuki; Matsuda, Tsuyoshi

2013-01-01

To improve magnetic resonance (MR) safety, we surveyed the accidents caused by large ferromagnetic materials brought into MR systems accidentally. We sent a questionnaire to 700 Japanese medical institutions and received 405 valid responses (58%). A total of 97 accidents in 77 institutions were observed and we analyzed them regarding incidental rate, the detail situation and environmental factors. The mean accident rate of each institute was 0.7/100,000 examinations, which was widely distributed (0-25.6/100,000) depending on the institute. In this survey, relatively small institutes with less than 500 beds tend to have these accidents more frequently (paccidents than those with less than 10 daily examinations. The institutes with 6-10 MR examinations daily have significantly more accidents than that with more than 10 daily MR examinations (paccidents were considered to be "prejudice" and "carelessness" but some advocate "ignorance." Though we could not find significant reduction in the institutes that have lectures and training for MR safety, we should continue lectures and training for MR safety to reduce accidents due to "ignorance."

Postulated accident conditions for air cleaning systems and radiological dose assessments for containment options

International Nuclear Information System (INIS)

Hilliard, R.K.; Postma, A.K.

1975-01-01

Ambient conditions and performance requirements for emergency air cleaning systems applicable to commercial LMFBR plants were studied. The focus of this study centered on aerosol removal under hypothetical core disruptive accident conditions. Effort completed includes a review of air cleaning systems related to LMFBR plants, selection of three reference containment system designs, postulation of the EACS design basis accident (EACS-DBA), analysis of thermal conditions resulting from the DBA, analysis of aerosol transport behavior following the DBA, and an estimate of bone dose at the site boundary for each of the reference plant designs. Reference plant concepts were a single containment system (e.g., FFTF), a double containment system (e.g., CRBRP with closed head compartment), and a containment-confinement design in which an inerted, sealed primary volume was located within a ventilated building whose exhaust was filtered. The reference design basis accident selected here involved release to the inner containment system of 1 percent of non-volatile solids and plutonium, 25 percent of core halogens, 25 percent of core volatile solids, 100 percent of core noble gases, 68 lbs of sodium vapor and 5000 lbs of liquid sodium. 13 references. (U.S.)

Retrieval system for emplaced spent unreprocessed fuel (SURF) in salt bed depository: accident event analysis and mechanical failure probabilities. Final report

International Nuclear Information System (INIS)

Bhaskaran, G.; McCleery, J.E.

1979-10-01

This report provides support in developing an accident prediction event tree diagram, with an analysis of the baseline design concept for the retrieval of emplaced spent unreprocessed fuel (SURF) contained in a degraded Canister. The report contains an evaluation check list, accident logic diagrams, accident event tables, fault trees/event trees and discussions of failure probabilities for the following subsystems as potential contributors to a failure: (a) Canister extraction, including the core and ram units; (b) Canister transfer at the hoist area; and (c) Canister hoisting. This report is the second volume of a series. It continues and expands upon the report Retrieval System for Emplaced Spent Unreprocessed Fuel (SURF) in Salt Bed Depository: Baseline Concept Criteria Specifications and Mechanical Failure Probabilities. This report draws upon the baseline conceptual specifications contained in the first report

Impact of the TMI accident on the French nuclear program and the safety analysis

International Nuclear Information System (INIS)

Fourest, B.; Boaretto, Y.; Cayol, A.; Droulers, Y.; Goudal, M.; Oury, J.M.

1980-04-01

Almost immediately after the TMI accident, Electricite de France (EdF), Framatome and the French safety authorities started a large scale program of actions designed to analyse and understand the causes of the accident, and draw lessons applicable in France. This paper discusses these actions and the main conclusions of TMI accident analysis in France, notably: the fundamental role of plant operators, and the importance of operator training, written instructions and procedures, and diagnostic aids; the importance of feeding back operating experience to design teams, and incorporating the results of accident and post-accident studies in operating procedures; the necessity to improve knowledge of core cooling modes, including during two-phase flow and natural circulation; measures to improve particular systems and components [fr

The development of a nuclear accident risk information system

International Nuclear Information System (INIS)

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

2001-01-01

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

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

Ferguson, D.R.; Deitrich, L.W.; Brown, N.W.; Waltar, A.E.

1978-01-01

Methods used for analysis of material behaviour, accident phenomenology and integrated accident calculations are reviewed. Applications of these methods to hypothetical LOF and TOP accidents are discussed. Recent results obtained from applications to FFTF and CRBRP are presented. (author)

Development of integrated accident management assessment technology

International Nuclear Information System (INIS)

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

2002-04-01

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

Application of activity theory to analysis of human-related accidents: Method and case studies

International Nuclear Information System (INIS)

Yoon, Young Sik; Ham, Dong-Han; Yoon, Wan Chul

2016-01-01

This study proposes a new approach to human-related accident analysis based on activity theory. Most of the existing methods seem to be insufficient for comprehensive analysis of human activity-related contextual aspects of accidents when investigating the causes of human errors. Additionally, they identify causal factors and their interrelationships with a weak theoretical basis. We argue that activity theory offers useful concepts and insights to supplement existing methods. The proposed approach gives holistic contextual backgrounds for understanding and diagnosing human-related accidents. It also helps identify and organise causal factors in a consistent, systematic way. Two case studies in Korean nuclear power plants are presented to demonstrate the applicability of the proposed method. Human Factors Analysis and Classification System (HFACS) was also applied to the case studies. The results of using HFACS were then compared with those of using the proposed method. These case studies showed that the proposed approach could produce a meaningful set of human activity-related contextual factors, which cannot easily be obtained by using existing methods. It can be especially effective when analysts think it is important to diagnose accident situations with human activity-related contextual factors derived from a theoretically sound model and to identify accident-related contextual factors systematically. - Highlights: • This study proposes a new method for analysing human-related accidents. • The method was developed based on activity theory. • The concept of activity system model and contradiction was used in the method. • Two case studies in nuclear power plants are presented. • The method is helpful to consider causal factors systematically and comprehensively.

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

Energy Technology Data Exchange (ETDEWEB)

PIEPHO, M.G.

1999-10-20

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

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

International Nuclear Information System (INIS)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul

2006-01-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

The TE coupled RELAP5/PANTHER/COBRA code package and methodology for integrated PWR accident analysis

Energy Technology Data Exchange (ETDEWEB)

Schneidesch, Christophe R.; Zhang, Jinzhao; Ammirabile, Luca; Dalleur, Jean-Paul [Suez-Tractebel Engineering, Avenue Ariane 7, B-1200 Brussels (Belgium)

2006-07-01

At Tractebel Engineering (TE), a dynamic coupling has been developed between the best estimate thermal hydraulics system code RELAP5 and the 3-dimensional neutronics code PANTHER via the transient analysis code linkage program TALINK. An interface between PANTHER and the subchannel core thermal-hydraulic analysis code COBRA 3C has been established for on-line calculation of the Departure from Nucleate Boiling Ratio (DNBR). In addition to the standard RELAP5-PANTHER coupling, the fully dynamic coupling of the RELAP5/PANTHER/COBRA3C-TE code package can be activated for evaluation purposes in which the PANTHER close-channel thermal-hydraulics module is replaced by the COBRA3C-TE with cross flow modelling and extended T-H flow conditions capabilities. The qualification of the RELAP5-PANTHER coupling demonstrated the robustness achieved by the combined 3-D neutron kinetics/system T-H code package for transient simulations. The coupled TE code package has been approved by the Belgian Safety Authorities and is used at TE for analyzing asymmetric PWR accidents with strong core-system interactions. In particular, the TE coupled code package was first used to develop a main steam line break in hot shutdown conditions (SLBHZP) accident analysis methodology based on the TE deterministic bounding approach. This methodology has been reviewed and accepted by the Belgian Safety Authorities for specific applications. Those specific applications are related to the power up-rate and steam generator replacement project of the Doel 2 plant or to the Tihange-3 SLB accident re-analysis. A coupled feedwater line break (FLB) accident analysis methodology is currently being reviewed for application approval. The results of coupled thermal-hydraulic and neutronic analysis of SLB and FLB show that there exist important margins in the traditional final safety analysis report (FSAR) accident analysis. Those margins can be used to increase the operational flexibility of the plants. Moreover, the

MAAP-CANDU simulations of LOCA/LOECI accidents at Darlington NGS

International Nuclear Information System (INIS)

Kwee, M.T.; Choi, M.H.; Leung, R.K.

1996-01-01

Severe accidents have been the subject of a great deal of analysis and research, particularly in the light water reactor community. Although severe accident analysis in Canada deuterium-uranium (CANDU) reactors has not been published abundantly, a significant body of research and analysis has been accumulated. This has occurred because CANDU has directly taken into consideration a set of severe accidents [e.g loss-of-coolant accidents (LOCAs) coincident with a loss-of-emergency-coolant injection (LOECI)] in the design basis. These accidents have served to define the design requirements that ensure the integrity of the heat transport system. The CANDU reactor design has inherent heat sinks such as the primary heat transport system, the secondary side, moderator system, and shielding system (shield tank and end shields). These heat sinks are significant and are able to moderate or terminate the progression of severe accidents that go beyond the design base cases. These types of accidents are typically analyzed at Ontario Hydro in conjunction with probabilistic safety analysis (PSA), where the severe accident consequences are analyzed by a series of conservative hand-calculation methods

Study on the code system for the off-site consequences assessment of severe nuclear accident

Energy Technology Data Exchange (ETDEWEB)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-12-15

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents.

Study on the code system for the off-site consequences assessment of severe nuclear accident

International Nuclear Information System (INIS)

Kim, Sora; Mn, Byung Il; Park, Ki Hyun; Yang, Byung Mo; Suh, Kyung Suk

2016-01-01

The importance of severe nuclear accidents and probabilistic safety assessment (PSA) were brought to international attention with the occurrence of severe nuclear accidents caused by the extreme natural disaster at Fukushima Daiichi nuclear power plant in Japan. In Korea, studies on level 3 PSA had made little progress until recently. The code systems of level 3 PSA, MACCS2 (MELCORE Accident Consequence Code System 2, US), COSYMA (COde SYstem from MAria, EU) and OSCAAR (Off-Site Consequence Analysis code for Atmospheric Releases in reactor accidents, JAPAN), were reviewed in this study, and the disadvantages and limitations of MACCS2 were also analyzed. Experts from Korea and abroad pointed out that the limitations of MACCS2 include the following: MACCS2 cannot simulate multi-unit accidents/release from spent fuel pools, and its atmospheric dispersion is based on a simple Gaussian plume model. Some of these limitations have been improved in the updated versions of MACCS2. The absence of a marine and aquatic dispersion model and the limited simulating range of food-chain and economic models are also important aspects that need to be improved. This paper is expected to be utilized as basic research material for developing a Korean code system for assessing off-site consequences of severe nuclear accidents

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

Science.gov (United States)

Nishite, Yoshiaki; Takesawa, Shingo

2016-01-01

Background: Accidents that occur during dialysis treatment are notified to the medical staff via alarms raised by the dialysis apparatus. Similar to such real accidents, apparatus activation or accidents can be reproduced by simulating a treatment situation. An alarm that corresponds to such accidents can be utilized in the simulation model. Objectives: The aim of this study was to create an extracorporeal circulation system (hereinafter, the circulation system) for dialysis machines so that it sets off five types of alarms for: 1) decreased arterial pressure, 2) increased arterial pressure, 3) decreased venous pressure, 4) increased venous pressure, and 5) blood leakage, according to the five types of accidents chosen based on their frequency of occurrence and the degree of severity. Materials and Methods: In order to verify the alarm from the dialysis apparatus connected to the circulation system and the accident corresponding to it, an evaluation of the alarm for its reproducibility of an accident was performed under normal treatment circumstances. The method involved testing whether the dialysis apparatus raised the desired alarm from the moment of control of the circulation system, and measuring the time it took until the desired alarm was activated. This was tested on five main models from four dialyzer manufacturers that are currently used in Japan. Results: The results of the tests demonstrated successful activation of the alarms by the dialysis apparatus, which were appropriate for each of the five types of accidents. The time between the control of the circulatory system to the alarm signal was as follows, 1) venous pressure lower limit alarm: 7 seconds; 2) venous pressure lower limit: 8 seconds; 3) venous pressure upper limit: 7 seconds; 4) venous pressure lower limit alarm: 2 seconds; and 5) blood leakage alarm: 19 seconds. All alarms were set off in under 20 seconds. Conclusions: Thus, we can conclude that a simulator system using an extracorporeal

Validation of the metal fuel version of the SAS4A accident analysis code

International Nuclear Information System (INIS)

Tentner, A.M.

1991-01-01

This paper describes recent work directed towards the validation of the metal fuel version of the SAS4A accident analysis code. The SAS4A code system has been developed at Argonne National Laboratory for the simulation of hypothetical severe accidents in Liquid Metal-Cooled Reactors (LMR), designed to operate in a fast neutron spectrum. SAS4A was initially developed for the analysis of oxide-fueled liquid metal-cooled reactors and has played an important role in the simulation and assessment of the energetics potential for postulated severe accidents in these reactors. Due to the current interest in the metal-fueled liquid metal-cooled reactors, a metal fuel version of the SAS4A accident analysis code is being developed in the Integral Fast Reactor program at Argonne. During such postulated accident scenarios as the unprotected (i.e. without scram) loss-of-flow and transient overpower events, a large number of interrelated physical phenomena occur during a relatively short time. These phenomena include transient heat transfer and hydrodynamic events, coolant boiling, and fuel and cladding melting and relocation. Due to strong neutronic feedbacks these events can significantly influence the reactor power history in the accident progression. The paper presents the results of a recent SAS4A simulation of the M7 TREAT experiment. 6 refs., 5 figs

The Analysis of the Contribution of Human Factors to the In-Flight Loss of Control Accidents

Science.gov (United States)

Ancel, Ersin; Shih, Ann T.

2012-01-01

In-flight loss of control (LOC) is currently the leading cause of fatal accidents based on various commercial aircraft accident statistics. As the Next Generation Air Transportation System (NextGen) emerges, new contributing factors leading to LOC are anticipated. The NASA Aviation Safety Program (AvSP), along with other aviation agencies and communities are actively developing safety products to mitigate the LOC risk. This paper discusses the approach used to construct a generic integrated LOC accident framework (LOCAF) model based on a detailed review of LOC accidents over the past two decades. The LOCAF model is comprised of causal factors from the domain of human factors, aircraft system component failures, and atmospheric environment. The multiple interdependent causal factors are expressed in an Object-Oriented Bayesian belief network. In addition to predicting the likelihood of LOC accident occurrence, the system-level integrated LOCAF model is able to evaluate the impact of new safety technology products developed in AvSP. This provides valuable information to decision makers in strategizing NASA's aviation safety technology portfolio. The focus of this paper is on the analysis of human causal factors in the model, including the contributions from flight crew and maintenance workers. The Human Factors Analysis and Classification System (HFACS) taxonomy was used to develop human related causal factors. The preliminary results from the baseline LOCAF model are also presented.

Severe accident tests and development of domestic severe accident system codes

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

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

Severe accident tests and development of domestic severe accident system codes

International Nuclear Information System (INIS)

2013-01-01

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

Detection of criticality accidents. The Intertechnique EDAC II system

International Nuclear Information System (INIS)

Prigent, R.

1991-01-01

The chief aim of the new generation of EDAC II criticality accidents detection system is to reduce the risks associated to the handling of fissile material by providing a swift and safe warning of the development of any criticality accident. To this function already devolving on the EDAC system of the previous generation, the EDAC II adds the possibility of storing in memory the characteristics of the accident, providing a daily follow-up of the striking events in the system through the print-out of a log book and providing assistance to the operators during the periodical tests. (Author)

Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

International Nuclear Information System (INIS)

Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

2000-01-01

Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

Review of current status for designing severe accident management support system

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwang Sub

2000-05-01

The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too.

Review of current status for designing severe accident management support system

International Nuclear Information System (INIS)

Jeong, Kwang Sub

2000-05-01

The development of operator support system (OSS) is ongoing in many other countries due to the complexity both in design and in operation for nuclear power plant. The computerized operator support system includes monitoring of some critical parameters, early detection of plant transient, monitoring of component status, plant maintenance, and safety parameter display, and the operator support system for these areas are developed and are being used in some plants. Up to now, the most operator support system covers the normal operation, abnormal operation, and emergency operation. Recently, however, the operator support system for severe accident is to be developed in some countries. The study for the phenomena of severe accident is not performed sufficiently, but, based on the result up to now, the operator support system even for severe accident will be developed in this study. To do this, at first, the current status of the operator support system for normal/abnormal/emergency operation is reviewed, and the positive aspects and negative aspects of systems are analyzed by their characteristics. And also, the major items that should be considered in designing the severe accident operator support system are derived from the review. With the survey of domestic and foreign operator support systems, they are reviewed in terms of the safety parameter display system, decision-making support system, and procedure-tracking system. For the severe accident, the severe accident management guideline (SAMG) which is developed by Westinghouse is reviewed; the characteristics, structure, and logical flow of SAMG are studied. In addition, the critical parameters for severe accident, which are the basis for operators decision-making in severe accident management and are supplied to the operators and the technical support center, are reviewed, too

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

OpenAIRE

Nishite, Yoshiaki; Takesawa, Shingo

2016-01-01

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

Analysis of an accident of local zone control system of 'pressure loss in the compartment water supply reservoir'

International Nuclear Information System (INIS)

Catana, A.

2001-01-01

This work presents the aftermath of a failure in the Zonal Control System caused by an accident of 'pressure loss in the compartment water supply reservoir' leading to an operational function fault of the liquid zonal control system. Causes for pressure drop may be several, as for instance: simultaneous mechanical fault of the three pumps, class IV total loss of power, a crack of reservoir, etc. Should this accident happens the reactor is shut down automatically by the digital control computer, on the 'ZONE CONTROL SYSTEM FAILURE' setback condition. The analyses were done hypothesizing that the covering gas system is functioning at design parameters and that the only possible accident is the one of pressure loss in supply reservoir. By making use of the software system developed at INR Pitesti, we could make the analysis of the phenomena which take place and thus we could obtain the evolution of the main parameters, namely, neutron and thermohydraulic parameters, as well as the actuating mode of the control and safety systems. Thus, by assuming a pressure drop under 8. 27 bar the 'SETBACK' system is triggered with a final value of the neutron power of 2% FP which can be reached with a power variation rate of 0.00086 decade/sec (- 0.1%/sec). In conclusion, the main parameters evolve as follows: 1. the water level in compartments is 'frozen' at a level at which the pressure in the supply reservoir is 7.3 bar; 2. the mechanical rods are gradually inserted, one bank first and a second one if necessary; 3. the shim rods are fully inserted; 4. the systems of SDS1 and SDS2 scram systems remain unactuated; 5. after 10 minutes from the 'SETBACK' triggering, the neutron power is reduced under 4%; 6. the thermohydraulic parameters of the primary circuit are maintained at normal values; 7. the thermohydraulic parameters of the secondary circuit are maintained at normal values

A Comparative analysis for control rod drop accident in WH and CE type

International Nuclear Information System (INIS)

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

2008-01-01

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

Source term analysis in severe accident induced by large break loss of coolant accident coincident with ship blackout for ship reactor

International Nuclear Information System (INIS)

Zhang Yanzhao; Zhang Fan; Zhao Xinwen; Zheng Yingfeng

2013-01-01

Using MELCOR code, the accident analysis model was established for a ship reactor. The behaviors of radioactive fission products were analyzed in the case of severe accident induced by large break loss of coolant accident coincident with ship blackout. The research mainly focused on the behaviors of release, transport, retention and the final distribution of inert gas and CsI. The results show that 83.12% of inert gas releases from the core, and the most of inert gas exists in the containment. About 83.08% of CsI release from the core, 72.66% of which is detained in the debris and the primary system, and 27.34% releases into the containment. The results can give a reference for the evaluation of cabin dose and nuclear emergency management. (authors)

Postulated accidents

International Nuclear Information System (INIS)

Ullrich, W.

1980-01-01

This lecture on 'Postulated Accidents' is the first of a series of lectures on the dynamic and transient behaviour of nuclear power plants, especially pressurized water reactors. The main points covered will be: Reactivity Accidents, Transients (Intact Loop) and Loss of Cooland Accidents (LOCA) including small leak. This lecture will discuss the accident analysis in general, the definition of the various operational phases, the accident classification, and, as an example, an accident sequence analysis on the basis of 'Postulated Accidents'. (orig./RW)

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

International Nuclear Information System (INIS)

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

1989-01-01

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

Thermal Hydraulic Analysis of K-DEMO Single Blanket Module for Preliminary Accident Analysis using MELCOR

Energy Technology Data Exchange (ETDEWEB)

Moon, Sung Bo; Bang, In Cheol [UNIST, Ulsan (Korea, Republic of)

2016-05-15

To develop the Korean fusion commercial reactor, preliminary design concept for K-DEMO (Korean fusion demonstration reactor) has been announced by NFRI (National Fusion Research Institute). This pre-conceptual study of K-DEMO has been introduced to identify technical details of a fusion power plant for the future commercialization of fusion reactor in Korea. Before this consideration, to build the K-DEMO, accident analysis is essential. Since the Fukushima accident, which is severe accident from unexpected disaster, safety analysis of nuclear power plant has become important. The safety analysis of both fission and fusion reactors is deemed crucial in demonstrating the low radiological effect of these reactors on the environment, during severe accidents. A risk analysis of K-DEMO should be performed, as a prerequisite for the construction of a fusion reactor. In this research, thermal-hydraulic analysis of single blanket module of K-DEMO is conducted for preliminary accident analysis for K-DEMO. Further study about effect of flow distributer is conducted. The normal K-DEMO operation condition is applied to the boundary condition and simulated to verify the material temperature limit using MELCOR. MELCOR is fully integrated, relatively fast-running code developed by Sandia National Laboratories. MELCOR had been used for Light Water Reactors and fusion reactor version of MELCOR was developed for ITER accident analysis. This study shows the result of thermal-hydraulic simulation of single blanket module with MELCOR which is severe accident code for nuclear fusion safety analysis. The difference of mass flow rate for each coolant channel with or without flow distributer is presented. With flow distributer, advantage of broadening temperature gradient in the K-DEMO blanket module and increase mass flow toward first wall is obtained. This can enhance the safety of K-DEMO blanket module. Most 13 .deg. C temperature difference in blanket module is obtained.

System 80+ design features for severe accident prevention and mitigation

International Nuclear Information System (INIS)

Jacob, M.C.; Schneider, R.E.; Finnicum, D.J.

1993-01-01

ABB-CE, in cooperation with the US Department of Energy, is working to develop and certify the System 80+ design, which is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the EPRI's Utility Requirements Document, and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the system is discussed along with its conformance to EPRI URD guidance, as applicable. Computer simulation of a best estimate severe accident scenario is presented to illustrate the acceptable containment performance of the design. It is concluded that by considering severe accident prevention and mitigation early in the design process, the System 80+ design represents a robust plant design that has low core damage frequencies, low containment conditional failure probabilities, and acceptable deterministic containment performance under severe accident conditions

Aerospace Accident - Injury Autopsy Data System -

Data.gov (United States)

Department of Transportation — The Aerospace Accident Injury Autopsy Database System will provide the Civil Aerospace Medical Institute (CAMI) Aerospace Medical Research Team (AMRT) the ability to...

Developement of integrated evaluation system for severe accident management

Energy Technology Data Exchange (ETDEWEB)

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

2005-04-01

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

Supporting system in emergency response plan for nuclear material transport accidents

International Nuclear Information System (INIS)

Nakagome, Y.; Aoki, S.

1993-01-01

As aiming to provide the detailed information concerning nuclear material transport accidents and to supply it to the concerned organizations by an online computer, the Emergency Response Supporting System has been constructed in the Nuclear Safety Technology Center, Japan. The system consists of four subsystems and four data bases. By inputting initial information such as name of package and date of accident, one can obtain the appropriate initial response procedures and related information for the accident immediately. The system must be useful for protecting the public safety from nuclear material transport accidents. But, it is not expected that the system shall be used in future. (J.P.N.)

Nuclear Reactor RA Safety Report, Vol. 13, Causes of possible accidents

International Nuclear Information System (INIS)

1986-11-01

This volume includes the analysis of possible accidents on the RA research reaktor. Any unwanted action causing decrease of integrity of any of the reactor safety barriers is considered to be a reactor accident. Safety barriers are: fuel element cladding, reactor vessel, biogical shield, and reactor building. Reactor accidents can be classified in four categories: (1) accidents caused by reactivity changes; (2) accidents caused by mis function of the cooling system; (3) accidents caused by errors in fuel management and auxiliary systems; (4) accidents caused by natural or other external disasters. The analysis of possible causes of reactor accidents includes the analysis of possible impacts on the reactor itself and the environment [sr

Thermal hydraulics of CANDU severe accident analysis

International Nuclear Information System (INIS)

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

2007-01-01

As interest in severe accident studies has increased in the last years, we have developed a set of simple models to analyze severe accidents in CANDU reactors that should be integrated in the EU codes. The CANDU600 reactor uses natural uranium fuel and heavy water (D 2 O) as both moderator and coolant, with the moderator and coolant in separate systems. We chose to analyze accident development for 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 temperatures inside a pressure tube reaches 1000 deg. C, a contact between pressure tube and calandria tube occurs and the residual 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) will be uncovered, then will disintegrate and fall down to the calandria vessel bottom. After all the quantity of 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 normally surrounds the calandria vessel. The phenomena described above are modelled, analyzed and compared with the available data. The results are encouraging. (authors)

Revisiting Ulchin 4 SGTR Accident - Analysis for EOP Improvement

Energy Technology Data Exchange (ETDEWEB)

Lee, Eun-Hye; Lee, Wook-Jo; Jerng, Dong-Wook [Chung-Ang University, Seoul (Korea, Republic of)

2016-10-15

The Steam Generator Tube Ruputure (SGTR) is an accident that U-tube inside the SG is defected so that the reactor coolant releases through broken U-tube and this is one of design basis accidents. Operating the Nuclear Power Plants (NPP), maintaing the integrity of core and preventing radiation release are most important things. Because of risks, many researchers have studied scenarios, impacts and the ways to mitigate SGTR accidents. The study to provide an experimental database of aerosol particle retention and to develop models to support accident management interventions during SGTR was performed. The scaled-down models of NPP were used for experiments, also, MELCOR and SCDAP/RELAP5 were used to simulate a design basis SGTR accident. This study had a major role to resolve uncertainties of various physical models for aerosol mechanical resuspension. The other study which analyzed SGTR accident for System-integrated Modular Advanced Reactor (SMART) was performed. In this analysis, the amount of break flow was focused and TASS/SMRS code was used. It assumed that maximum leak was generated, and found that high RCS pressure, low core inlet coolant temperature, and low break location of the SG cassette contributed to leakage. Although the leakage was large, there was no direct release to atmosphere because the pressure of secondary loop was maintained below the safety relief valve set point. In this analysis, comparison of mitigating procedure when SGTR occurs between shutdown condition and full power condition was performed. In shutdown condition, the core uncovery would not take place in 16 hours whether the cooling procedures are performed or not. Therefore, the integrated amount of break flow should be considered only. In this point of view, cooling through intact SG only, case 3, is the best way to minimize the amount of break flow. In full power condition, the core water level is changed due to high reactor power. The important thing to protect NPP is to keep

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

This document provides the detailed accident analysis to support HNF-3553, Annex B, Spent Nuclear Fuel Project Final Safety Analysis Report (FSAR), ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' All assumptions, parameters, and models used to provide the analysis of the design basis accidents are documented to support the conclusions in the FSAR. The calculations in this document address the design basis accidents (DBAs) selected for analysis in HNF-3553, ''Spent Nuclear Fuel Project Final Safety Analysis Report'', Annex B, ''Cold Vacuum Drying Facility Final Safety Analysis Report.'' The objective is to determine the quantity of radioactive particulate available for release at any point during processing at the Cold Vacuum Drying Facility (CVDF) and to use that quantity to determine the amount of radioactive material released during the DBAs. The radioactive material released is used to determine dose consequences to receptors at four locations, and the dose consequences are compared with the appropriate evaluation guidelines and release limits to ascertain the need for preventive and mitigative controls.

Development of Information Display System for Operator Support in Severe Accident

International Nuclear Information System (INIS)

Jeong, Kwang Il; Lee, Joon Ku

2016-01-01

When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future

Post-accident monitoring systems in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Suriya Murthy, N.; Sivasailanathan, Vidhya; Ananth, Allu; Roy, Kallol

2018-01-01

PFBR is a 500 MW(e) MOX fueled and sodium cooled fast reactor (SFR) under advanced stage of commissioning at Kalpakkam. Currently, the main vessel is preheated and sodium has been charged into two secondary loops that are operated in recirculation mode. In order to characterize the radiation field and contamination, the workplace monitoring is undertaken using installed monitors that are commissioned and made operational. This helps to ensure radiological protection during normal operating conditions. On the other hand, radiological monitoring in emergency conditions is quite different. For undertaking the mitigative accident management, a set of specialized nuclear instruments called post-accident monitoring systems (PAMS) which include radiation monitors are stipulated. The Fukushima Daiichi accident emphasized the importance and need for reliable accident monitoring instrumentation to indicate the safety functions during the progression and aftermath of accident in NPP. In PFBR, the PAMS are integrated with other monitoring systems in design stage itself to manage the measurements and indicating the safety functions for implementing EOP and SAMG

Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

International Nuclear Information System (INIS)

Rao, Suman

2007-01-01

One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly

Design and implementation of an identification system in construction site safety for proactive accident prevention.

Science.gov (United States)

Yang, Huanjia; Chew, David A S; Wu, Weiwei; Zhou, Zhipeng; Li, Qiming

2012-09-01

Identifying accident precursors using real-time identity information has great potential to improve safety performance in construction industry, which is still suffering from day to day records of accident fatality and injury. Based on the requirements analysis for identifying precursor and the discussion of enabling technology solutions for acquiring and sharing real-time automatic identification information on construction site, this paper proposes an identification system design for proactive accident prevention to improve construction site safety. Firstly, a case study is conducted to analyze the automatic identification requirements for identifying accident precursors in construction site. Results show that it mainly consists of three aspects, namely access control, training and inspection information and operation authority. The system is then designed to fulfill these requirements based on ZigBee enabled wireless sensor network (WSN), radio frequency identification (RFID) technology and an integrated ZigBee RFID sensor network structure. At the same time, an information database is also designed and implemented, which includes 15 tables, 54 queries and several reports and forms. In the end, a demonstration system based on the proposed system design is developed as a proof of concept prototype. The contributions of this study include the requirement analysis and technical design of a real-time identity information tracking solution for proactive accident prevention on construction sites. The technical solution proposed in this paper has a significant importance in improving safety performance on construction sites. Moreover, this study can serve as a reference design for future system integrations where more functions, such as environment monitoring and location tracking, can be added. Copyright © 2011 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

Hill, Axel; Stiepani, Cristoph; Drechsler, Michael

2015-01-01

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

Methodology for time-dependent reliability analysis of accident sequences and complex reactor systems

International Nuclear Information System (INIS)

Paula, H.M.

1984-01-01

The work presented here is of direct use in probabilistic risk assessment (PRA) and is of value to utilities as well as the Nuclear Regulatory Commission (NRC). Specifically, this report presents a methodology and a computer program to calculate the expected number of occurrences for each accident sequence in an event tree. The methodology evaluates the time-dependent (instantaneous) and the average behavior of the accident sequence. The methodology accounts for standby safety system and component failures that occur (a) before they are demanded, (b) upon demand, and (c) during the mission (system operation). With respect to failures that occur during the mission, this methodology is unique in the sense that it models components that can be repaired during the mission. The expected number of system failures during the mission provides an upper bound for the probability of a system failure to run - the mission unreliability. The basic event modeling includes components that are continuously monitored, periodically tested, and those that are not tested or are otherwise nonrepairable. The computer program ASA allows practical applications of the method developed. This work represents a required extension of the presently available methodology and allows a more realistic PRA of nuclear power plants

Aspects of severe accidents in transmutation systems

International Nuclear Information System (INIS)

Wider, H.U.; Karlson, J.; Jones, A.V.

2001-01-01

The different types of transmutation systems under investigation include accelerator driven (ADS) and critical systems. To switch off an accelerator in case of an accident initiation is quite important for all accidents. For a fast ADS the grace times available for doing so depend strongly on the total heat capacity and the natural circulation capability of the primary coolant. Cooling with heavy metal Pb-Bi has considerable advantages in this regard compared to gas cooling. Moreover it allows passive ex-vessel cooling with natural air or water circulation. In the remote likelihood of fuel melting, oxide fuel appears to mix with the Pb-Bi coolant. Fast critical systems that are cooled by Pb-Bi will automatically shut off if the flow or heat sink is lost. Reactivity accidents can be limited by a low total control rod worth. High temperature reactors can achieve only incomplete burning of actinides. If an accelerator is added to increase burn-up, a fast spectrum region is needed, which has a low heat capacity. (author)

Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach.

Science.gov (United States)

Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

2017-06-01

Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach's alpha of 75%. Data was analyzed using the decision Delphi technique. GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information.

Statistical analysis of the early phase of SBO accident for PWR

Energy Technology Data Exchange (ETDEWEB)

Kozmenkov, Yaroslav, E-mail: y.kozmenkov@hzdr.de; Jobst, Matthias, E-mail: m.jobst@hzdr.de; Kliem, Soeren, E-mail: s.kliem@hzdr.de; Schaefer, Frank, E-mail: f.schaefer@hzdr.de; Wilhelm, Polina, E-mail: p.wilhelm@hzdr.de

2017-04-01

Highlights: • Best estimate model of generic German PWR is used in ATHLET-CD simulations. • Uncertainty and sensitivity analysis of the early phase of SBO accident is presented. • Prediction intervals for occurrence of main events are evaluated. - Abstract: A statistical approach is used to analyse the early phase of station blackout accident for generic German PWR with the best estimate system code ATHLET-CD as a computation tool. The analysis is mainly focused on the timescale uncertainties of the accident events which can be detected at the plant. The developed input deck allows variations of all input uncertainty parameters relevant to the case. The list of identified and quantified input uncertainties includes 30 parameters related to the simulated physical phenomena/processes. Time uncertainties of main events as well as the major contributors to these uncertainties are defined. The uncertainty in decay heat has the highest contribution to the uncertainties of the analysed events. A linear regression analysis is used for predicting times of future events from detected times of occurred/past events. An accuracy of event predictions is estimated and verified. The presented statistical approach could be helpful for assessing and improving existing or elaborating additional emergency operating procedures aimed to prevent severe damage of reactor core.

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

International Nuclear Information System (INIS)

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

2005-03-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2005-03-15

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

Utilization of accident databases and fuzzy sets to estimate frequency of HazMat transport accidents

International Nuclear Information System (INIS)

Qiao Yuanhua; Keren, Nir; Mannan, M. Sam

2009-01-01

Risk assessment and management of transportation of hazardous materials (HazMat) require the estimation of accident frequency. This paper presents a methodology to estimate hazardous materials transportation accident frequency by utilizing publicly available databases and expert knowledge. The estimation process addresses route-dependent and route-independent variables. Negative binomial regression is applied to an analysis of the Department of Public Safety (DPS) accident database to derive basic accident frequency as a function of route-dependent variables, while the effects of route-independent variables are modeled by fuzzy logic. The integrated methodology provides the basis for an overall transportation risk analysis, which can be used later to develop a decision support system.

Analysis of Three Mile Island-Unit 2 accident

International Nuclear Information System (INIS)

1980-03-01

The Nuclear Safety Analysis Center (NSAC) of the Electric Power Research Institute has analyzed the Three Mile Island-2 accident. Early results of this analysis were a brief narrative summary, issued in mid-May 1979 and an initial version of this report issued later in 1979 as noted in the Foreword. The present report is a revised version of the 1979 report, containing summaries, a highly detailed sequence of events, a comparison of that sequence of events with those from other sources, 25 appendices, references and a list of abbreviations and acronyms. A matrix of equipment and system actions is included as a folded insert

Accident tolerant high-pressure helium injection system concept for light water reactors

International Nuclear Information System (INIS)

Massey, Caleb; Miller, James; Vasudevamurthy, Gokul

2016-01-01

Highlights: • Potential helium injection strategy is proposed for LWR accident scenarios. • Multiple injection sites are proposed for current LWR designs. • Proof-of-concept experimentation illustrates potential helium injection benefits. • Computational studies show an increase in pressure vessel blowdown time. • Current LOCA codes have the capability to include helium for feasibility calculations. - Abstract: While the design of advanced accident-tolerant fuels and structural materials continues to remain the primary focus of much research and development pertaining to the integrity of nuclear systems, there is a need for a more immediate, simple, and practical improvement in the severe accident response of current emergency core cooling systems. Current blowdown and reflood methodologies under accident conditions still allow peak cladding temperatures to approach design limits and detrimentally affect the integrity of core components. A high-pressure helium injection concept is presented to enhance accident tolerance by increasing operator response time while maintaining lower peak cladding temperatures under design basis and beyond design basis scenarios. Multiple injection sites are proposed that can be adapted to current light water reactor designs to minimize the need for new infrastructure, and concept feasibility has been investigated through a combination of proof-of-concept experimentation and computational modeling. Proof-of-concept experiments show promising cooling potential using a high-pressure helium injection concept, while the developed choked-flow model shows core depressurization changes with added helium injection. Though the high-pressure helium injection concept shows promise, future research into the evaluation of system feasibility and economics are needed.Classification: L. Safety and risk analysis

Analysis of accidents in nine Iranian gas refineries: 2007-2011.

Science.gov (United States)

Mehrdad, R; Bolouri, A; Shakibmanesh, A R

2013-10-01

Occupational accidents are one of the major health hazards in industries and associated with high mortality, morbidity, spiritual damage and economic losses in the world. To determine the incidence of occupational accidents in 9 Iranian gas refineries between March 2007 and February 2011. Data on all occupational accidents occurred between March 2007 and February 2011, as well as other possible associated variables including time of accident, whether the accident was due to a personal or systemic fault, type of accident and its outcomes, age and gender of the victim, the injured parts of the body, job experience, and type of employment, were extracted from HSE reports and notes of health care services. Based on these data, we calculated the incidence rate of accidents and assessed the associated factors. During the 5 studied years, 1129 accidents have been recorded. The incidence of fatal accidents was 1.64 per 100 000 and of nonfatal accidents was 1857 per 100 000 workers per year. 99.4% of injured workers were male. The mean±SD age of injured people was 29.6±7.3 years. Almost 70% of injured workers aged under 30 years. The mean±SD job experience was 5.3±5.3 years. Accidents occurred more commonly around 10:00. More than 60% of accidents happened between 8:00 and 15:00. July had the highest incidence rate. The most common type of accident was being struck by an object (48%). More than 94% of accidents are caused by personal rather than systemic faults. Hands and wrists were the most common injured parts and involved in more than one-third of accidents. 70% of injured workers needed medical treatment and returned to work after primary treatment. The pattern of occupational accidents in Iranian gas refineries is similar to other previous reports in many ways. The incidence did not change significantly over the study period. Establishment of an online network for precise registration, notification and meticulous data collection seems necessary.

Development of Information Display System for Operator Support in Severe Accident

Energy Technology Data Exchange (ETDEWEB)

Jeong, Kwang Il; Lee, Joon Ku [KAERI, Daejeon (Korea, Republic of)

2016-05-15

When the severe accident occurs, the technical support center (TSC) performs the mitigation strategy with severe accident management guidelines (SAMG) and communicates with main control room (MCR) operators to obtain information of plant's status. In such circumstances, the importance of an information display for severe accident is increased. Therefore an information display system dedicated to severe accident conditions is required to secure the plant information, to provide the necessary information to MCR operators and TSC operators, and to support the decision using these information. We setup the design concept of severe accident information display system (SIDS) in the previous study and defined its requirements of function and performance. This paper describes the process, results of the identification of the severe accident information for MCR operator and the implementation of SIDS. Further implementation on post-accident monitoring function and data validation function for severe accidents will be accomplished in the future.

Control assembly ejection accident analysis for WWER-440 (Armenian NPP)

International Nuclear Information System (INIS)

Bznuni, S.; Malakyan, Ts.; Amirjanyan, A.; Ghasabyan, L.

2007-01-01

Control Assembly ejection in WWER-440 initiated by the loss of integrity of the Control Assemblies drive housing has been analyzed. This event causes a very rapid reactivity insertion to the core and small break LOCA which potentially could lead to rapid power increase and redistribution of heat release in the core resulting in a fuel, cladding and coolant temperature rise; primary pressure increase, radiological consequences due to loss of primary coolant and potential loss of cladding integrity and fuel disintegration (if applicable). Methodology of the analysis is based on conservative assumptions as well as on deterministic approach for selection of functioning logic of systems and equipment's to maximize reactor core power and minimize power decreasing reactivity feedback. Computational analyses were performed by 3D kinetics PARCS-RELAP coupled code. WWER-440 fuel cross-section libraries, diffusion coefficients and kinetics parameters were calculated by HELOS code. In this paper analysis of accident for Hot Full Power was presented. Results of analysis show that ANPP WWER-440 reactor design meets acceptance criteria prescribed for RIA type design based accidents (Authors)

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

International Nuclear Information System (INIS)

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

2013-01-01

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

RA reactor safety analysis, Part II - Accident analysis; Analiza sigurnosti rada Reaktora RA I-III, Deo II - Analiza akcidenta

Energy Technology Data Exchange (ETDEWEB)

Raisic, N; Radanovic, Lj; Milovanovic, M; Afgan, N; Kulundzic, P [Institute of Nuclear Sciences Boris Kidric, Vinca, Beograd (Serbia and Montenegro)

1963-02-15

This part of the RA reactor safety analysis includes analysis of possible accidents caused by failures of the reactor devices and errors during reactor operation. Two types of accidents are analyzed: accidents resulting from uncontrolled reactivity increase, and accidents caused by interruption of cooling.

Internal event analysis for Laguna Verde Unit 1 Nuclear Power Plant. Accident sequence quantification and results

International Nuclear Information System (INIS)

Huerta B, A.; Aguilar T, O.; Nunez C, A.; Lopez M, R.

1994-01-01

The Level 1 results of Laguna Verde Nuclear Power Plant PRA are presented in the I nternal Event Analysis for Laguna Verde Unit 1 Nuclear Power Plant, CNSNS-TR 004, in five volumes. The reports are organized as follows: CNSNS-TR 004 Volume 1: Introduction and Methodology. CNSNS-TR4 Volume 2: Initiating Event and Accident Sequences. CNSNS-TR 004 Volume 3: System Analysis. CNSNS-TR 004 Volume 4: Accident Sequence Quantification and Results. CNSNS-TR 005 Volume 5: Appendices A, B and C. This volume presents the development of the dependent failure analysis, the treatment of the support system dependencies, the identification of the shared-components dependencies, and the treatment of the common cause failure. It is also presented the identification of the main human actions considered along with the possible recovery actions included. The development of the data base and the assumptions and limitations in the data base are also described in this volume. The accident sequences quantification process and the resolution of the core vulnerable sequences are presented. In this volume, the source and treatment of uncertainties associated with failure rates, component unavailabilities, initiating event frequencies, and human error probabilities are also presented. Finally, the main results and conclusions for the Internal Event Analysis for Laguna Verde Nuclear Power Plant are presented. The total core damage frequency calculated is 9.03x 10-5 per year for internal events. The most dominant accident sequences found are the transients involving the loss of offsite power, the station blackout accidents, and the anticipated transients without SCRAM (ATWS). (Author)

The role of quantitative uncertainty in the safety analysis of flammable gas accidents in Hanford waste tanks

International Nuclear Information System (INIS)

Bratzel, D.R.

1998-01-01

Following a 1990 investigation into flammable gas generation, retention, and release mechanisms within the Hanford Site high-level waste tanks, personnel concluded that the existing Authorization Basis documentation did not adequately evaluate flammable gas hazards. The US Department of Energy Headquarters subsequently declared the flammable gas hazard as an unresolved safety issue. Although work scope has been focused on resolution of the issue, it has yet to be resolved due to considerable uncertainty regarding essential technical parameters and associated risk. Resolution of the Flammable Gas Safety Issue will include the identification of a set of controls for the Authorization Basis for the tanks which will require a safety analysis of flammable gas accidents. A traditional nuclear facility safety analysis is based primarily on the analysis of a set of bounding accidents to represent the risks of the possible accidents and hazardous conditions at a facility. While this approach may provide some indication of the bounding consequences of accidents for facilities, it does not provide a satisfactory basis for identification of facility risk or safety controls when there is considerable uncertainty associated with accident phenomena and/or data as is the case with potential flammable gas accidents at the Hanford Site. This is due to the difficulties in identifying the bounding case and reaching consensus among safety analysts, facility operations and engineering, and the regulator on the implications of the safety analysis results. In addition, the bounding cases are frequently based on simplifying assumptions that make the analysis results insensitive to variations among facilities or the impact of alternative safety control strategies. The existing safety analysis of flammable gas accidents for the Tank Waste Remediation system (TWRS) at the Hanford Site has these difficulties. However, Hanford Site personnel are developing a refined safety analysis approach

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

Energy Technology Data Exchange (ETDEWEB)

Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong [Korea Power Engineering Company, Inc, 150 Deokjin-dong, Yuseong-gu, Daejeon, 305-353 (Korea, Republic of)

2006-07-01

This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

A coupled 3-D kinetics/system thermal-hydraulic analysis of main steam line break accident for Optimized Power Reactor 1000

International Nuclear Information System (INIS)

Jin, Yung Kwon; Choi, Chul Jin; Kim, Eun Kee; Lee, Sang Yong

2006-01-01

This paper presents the results of the coupled 3-D neutronics/thermal-hydraulic analysis of hypothetical main steam line break (MSLB) accident for Optimized Power Reactor 1000. One of the major concerns of this accident is a return-to-power occurrence accompanied with extremely large radial peaking near the stuck Control Element Assembly (CEA). The conventional point kinetics application does not properly account for this kind of asymmetric and local core behavior. Therefore, the current licensing method of point kinetics application introduces some uncertainties and conservatisms in the physics parameters generation, e.g., the static net scram rod worth, moderator cooldown reactivity, Doppler reactivity, and a 3-D peaking factor. The recently developed UNICORN-TM code system is applied for the 3-D coupled calculation, where neutronics code MASTER is coupled with the best-estimate system transient code RETRAN. The 3-D coupled results were assessed in comparison with those by point kinetics application using stand-alone RETRAN application. To quantify the 3-D reactivity benefits over point kinetics, both calculations assumed the accidents to be initiated from the same core state, e.g., end of cycle burnup, fuel and CEA configuration with the same initial moderator and Doppler temperature coefficient, and with initial system thermal-hydraulic condition. The core physics parameters required for point kinetics application were produced using MASTER with the method and procedure consistent with the current licensing application. The occurrence of return-to-power was simulated by intentionally reducing the net CEA worth in order to assess the spatial power distribution and local T-H effect on the dynamic reactivity feedback. The results have demonstrated that the 3-D analysis removes some of the conservatisms inherent in point kinetics analysis mainly caused by the inability to properly account for local reactivity feedback effects during return-to-power transient

10 CFR 835.1304 - Nuclear accident dosimetry.

Science.gov (United States)

2010-01-01

...) Methods and equipment for analysis of biological materials; (3) A system of fixed nuclear accident... Nuclear accident dosimetry. (a) Installations possessing sufficient quantities of fissile material to... 10 Energy 4 2010-01-01 2010-01-01 false Nuclear accident dosimetry. 835.1304 Section 835.1304...

Passive depressurization accident management strategy for boiling water reactors

International Nuclear Information System (INIS)

Liu, Maolong; Erkan, Nejdet; Ishiwatari, Yuki; Okamoto, Koji

2015-01-01

Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident

Passive depressurization accident management strategy for boiling water reactors

Energy Technology Data Exchange (ETDEWEB)

Liu, Maolong, E-mail: liuml@vis.t.u-tokyo.ac.jp [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Erkan, Nejdet [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan); Ishiwatari, Yuki [Department of Nuclear Engineering and Management, School of Engineering, The University of Tokyo (Japan); Hitachi-GE Nuclear Energy, Ltd. (Japan); Okamoto, Koji [Nuclear Professional School, School of Engineering, The University of Tokyo (Japan)

2015-04-01

Highlights: • We proposed two passive depressurization systems for BWR severe accident management. • Sensitivity analysis of the passive depressurization systems with different leakage area. • Passive depressurization strategies can prevent direct containment heating. - Abstract: According to the current severe accident management guidance, operators are required to depressurize the reactor coolant system to prevent or mitigate the effects of direct containment heating using the safety/relief valves. During the course of a severe accident, the pressure boundary might fail prematurely, resulting in a rapid depressurization of the reactor cooling system before the startup of SRV operation. In this study, we demonstrated that a passive depressurization system could be used as a severe accident management tool under the severe accident conditions to depressurize the reactor coolant system and to prevent an additional devastating sequence of events and direct containment heating. The sensitivity analysis performed with SAMPSON code also demonstrated that the passive depressurization system with an optimized leakage area and failure condition is more efficient in managing a severe accident.

[An analysis of industrial accidents in the working field with a particular emphasis on repeated accidents].

Science.gov (United States)

Wakisaka, I; Yanagihashi, T; Tomari, T; Sato, M

1990-03-01

The present study is based on an analysis of routinely submitted reports of occupational accidents experienced by the workers of industrial enterprises under the jurisdiction of Kagoshima Labor Standard Office during a 5-year period 1983 to 1987. Officially notified injuries serious enough to keep employees away from their job for work at least 4 days were utilized in this study. Data was classified so as to give an observed frequency distribution for workers having any specified number of accidents. Also, the accident rate which is an indicator of the risk of accident was compared among different occupations, between age groups and between the sexes. Results obtained are as follows; 1) For the combined total of 6,324 accident cases for 8 types of occupation (Construction, Transportation, Mining & Quarrying, Forestry, Food manufacture, Lumber & Woodcraft, Manufacturing industry and Other business), the number of those who had at least one accident was 6,098, of which 5,837 were injured only once, 208 twice, 21 three times and 2 four times. When occupation type was fixed, however, the number of workers having one, two, three and four times of accidents were 5,895, 182, 19 and 2, respectively. This suggests that some workers are likely to have experienced repeated accidents in more than one type of occupation.(ABSTRACT TRUNCATED AT 250 WORDS)

Anthropotechnological analysis of industrial accidents in Brazil.

Science.gov (United States)

Binder, M. C.; de Almeida, I. M.; Monteau, M.

1999-01-01

The Brazilian Ministry of Labour has been attempting to modify the norms used to analyse industrial accidents in the country. For this purpose, in 1994 it tried to make compulsory use of the causal tree approach to accident analysis, an approach developed in France during the 1970s, without having previously determined whether it is suitable for use under the industrial safety conditions that prevail in most Brazilian firms. In addition, opposition from Brazilian employers has blocked the proposed changes to the norms. The present study employed anthropotechnology to analyse experimental application of the causal tree method to work-related accidents in industrial firms in the region of Botucatu, São Paulo. Three work-related accidents were examined in three industrial firms representative of local, national and multinational companies. On the basis of the accidents analysed in this study, the rationale for the use of the causal tree method in Brazil can be summarized for each type of firm as follows: the method is redundant if there is a predominance of the type of risk whose elimination or neutralization requires adoption of conventional industrial safety measures (firm representative of local enterprises); the method is worth while if the company's specific technical risks have already largely been eliminated (firm representative of national enterprises); and the method is particularly appropriate if the firm has a good safety record and the causes of accidents are primarily related to industrial organization and management (multinational enterprise). PMID:10680249

Safety culture and accident analysis-A socio-management approach based on organizational safety social capital

Energy Technology Data Exchange (ETDEWEB)

Rao, Suman [Risk Analyst (India)]. E-mail: sumanashokrao@yahoo.co.in

2007-04-11

One of the biggest challenges for organizations in today's competitive business environment is to create and preserve a self-sustaining safety culture. Typically, Key drivers of safety culture in many organizations are regulation, audits, safety training, various types of employee exhortations to comply with safety norms, etc. However, less evident factors like networking relationships and social trust amongst employees, as also extended networking relationships and social trust of organizations with external stakeholders like government, suppliers, regulators, etc., which constitute the safety social capital in the Organization-seem to also influence the sustenance of organizational safety culture. Can erosion in safety social capital cause deterioration in safety culture and contribute to accidents? If so, how does it contribute? As existing accident analysis models do not provide answers to these questions, CAMSoC (Curtailing Accidents by Managing Social Capital), an accident analysis model, is proposed. As an illustration, five accidents: Bhopal (India), Hyatt Regency (USA), Tenerife (Canary Islands), Westray (Canada) and Exxon Valdez (USA) have been analyzed using CAMSoC. This limited cross-industry analysis provides two key socio-management insights: the biggest source of motivation that causes deviant behavior leading to accidents is 'Faulty Value Systems'. The second biggest source is 'Enforceable Trust'. From a management control perspective, deterioration in safety culture and resultant accidents is more due to the 'action controls' rather than explicit 'cultural controls'. Future research directions to enhance the model's utility through layering are addressed briefly.

Design and Development of a Severe Accident Training System

International Nuclear Information System (INIS)

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

2005-01-01

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

Sensitivity and uncertainty analysis for Ignalina NPP confinement in case of loss of coolant accident

International Nuclear Information System (INIS)

Urbonavicius, E.; Babilas, E.; Rimkevicius, S.

2003-01-01

At present the best-estimate approach in the safety analysis of nuclear power plants is widely used around the world. The application of such approach requires to estimate the uncertainty of the calculated results. Various methodologies are applied in order to determine the uncertainty with the required accuracy. One of them is the statistical methodology developed at GRS mbH in Germany and integrated into the SUSA tool, which was applied for the sensitivity and uncertainty analysis of the thermal-hydraulic parameters inside the confinement (Accident Localisation System) of Ignalina NPP with RBMK-1500 reactor in case of Maximum Design Basis Accident (break of 900 mm diameter pipe). Several parameters that could potentially influence the calculated results were selected for the analysis. A set of input data with different initial values of the selected parameters was generated. In order to receive the results with 95 % probability and 95 % accuracy, 100 runs were performed with COCOSYS code developed at GRS mbH. The calculated results were processed with SUSA tool. The performed analysis showed a rather low dispersion of the results and only in the initial period of the accident. Besides, the analysis showed that there is no threat to the building structures of Ignalina NPP confinement in case of the considered accident scenario. (author)

Feature article. Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

Ekarinai, Masashi; Ake, Yutaka; Narabayashi, Tadashi

2011-01-01

This special feature article consisted of five reports and the minutes of emergency discussion meeting on Fukushima Daiichi Nuclear Power Plant (NPP) accident. Effects of the accident on future electricity supply of electric utilities and also on business development of nuclear industries were discussed. Activities of senior network team of atomic energy society of Japan (AESJ) to conduct severe accident analysis and early restoration from the accident were introduced. Circulating injection reactor cooling system and zeolite decontamination system of accumulated contaminated water was proposed. Effects of the accident on overseas reaction on nuclear development were also reported as well as personal experience of the professor in the US west coast on communications. (T. Tanaka)

MELCOR DB Construction for the Severe Accident Analysis DB

International Nuclear Information System (INIS)

Song, Y. M.; Ahn, K. I.

2011-01-01

The Korea Atomic Energy Research Institute (KAERI) has been constructing a severe accident analysis database (DB) under a National Nuclear R and D Program. In particular, an MAAP (commercial code being widely used in industries for integrated severe accident analysis) DB for many scenarios including a station blackout (SBO) has been completed. This paper shows the MELCOR DB construction process with examples of SBO scenarios, and the results will be used for a comparison with the MAAP DB

Prediction accident triangle in maintenance of underground mine facilities using Poisson distribution analysis

Science.gov (United States)

Khuluqi, M. H.; Prapdito, R. R.; Sambodo, F. P.

2018-04-01

In Indonesia, mining is categorized as a hazardous industry. In recent years, a dramatic increase of mining equipment and technological complexities had resulted in higher maintenance expectations that accompanied by the changes in the working conditions, especially on safety. Ensuring safety during the process of conducting maintenance works in underground mine is important as an integral part of accident prevention programs. Accident triangle has provided a support to safety practitioner to draw a road map in preventing accidents. Poisson distribution is appropriate for the analysis of accidents at a specific site in a given time period. Based on the analysis of accident statistics in the underground mine maintenance of PT. Freeport Indonesia from 2011 through 2016, it is found that 12 minor accidents for 1 major accident and 66 equipment damages for 1 major accident as a new value of accident triangle. The result can be used for the future need for improving the accident prevention programs.

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

Energy Technology Data Exchange (ETDEWEB)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.).

Loss-of-coolant and loss-of-flow accident in the ITER-EDA first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-05-01

This report presents the analysis of the transient thermal-hydraulic system behaviour inside the first wall/blanket cooling system and the resulting temperature response inside the first wall and blanket of the ITER-EDA (International Thermonuclear Experimental Reactor - Engineering Design Activities) reactor design during a: - Loss-of-coolant accident caused by a reputure of the pump suction pipe; - loss-of-flow accident caused by a trip of the recirculation pump. (orig.)

Geographic Information System (GIS) capabilities in traffic accident information management: a qualitative approach

Science.gov (United States)

Ahmadi, Maryam; Valinejadi, Ali; Goodarzi, Afshin; Safari, Ameneh; Hemmat, Morteza; Majdabadi, Hesamedin Askari; Mohammadi, Ali

2017-01-01

Background Traffic accidents are one of the more important national and international issues, and their consequences are important for the political, economical, and social level in a country. Management of traffic accident information requires information systems with analytical and accessibility capabilities to spatial and descriptive data. Objective The aim of this study was to determine the capabilities of a Geographic Information System (GIS) in management of traffic accident information. Methods This qualitative cross-sectional study was performed in 2016. In the first step, GIS capabilities were identified via literature retrieved from the Internet and based on the included criteria. Review of the literature was performed until data saturation was reached; a form was used to extract the capabilities. In the second step, study population were hospital managers, police, emergency, statisticians, and IT experts in trauma, emergency and police centers. Sampling was purposive. Data was collected using a questionnaire based on the first step data; validity and reliability were determined by content validity and Cronbach’s alpha of 75%. Data was analyzed using the decision Delphi technique. Results GIS capabilities were identified in ten categories and 64 sub-categories. Import and process of spatial and descriptive data and so, analysis of this data were the most important capabilities of GIS in traffic accident information management. Conclusion Storing and retrieving of descriptive and spatial data, providing statistical analysis in table, chart and zoning format, management of bad structure issues, determining the cost effectiveness of the decisions and prioritizing their implementation were the most important capabilities of GIS which can be efficient in the management of traffic accident information. PMID:28848627

Research on sever accident emergency simulation system for CPR1000

International Nuclear Information System (INIS)

Yang Zhifei; Liao Yehong; Liang Manchun; Li Ke; Yang Jie; Chen Yali

2015-01-01

The enhanced capability to nuclear power plant (NPP) severe accident management and emergency response depends heavily on exercises. Since the exercise scene is usually monotonous and not realistic, and conduct of exercise has a high cost, the effect of enhancing the capability is limited. Thus, the development of a Sever Accident Emergency Simulation System (SAESS) is necessary. SAESS is able to connect NPP simulator, and simulates the process of severe accident management, personnel evacuation, the dispersion of radioactive plume, and emergency response of emergency organizations. The system helps to design several of exercise scenes and optimize the disposal strategy in different severe accidents. In addition, the system reduces the cost of emergency exercise by computer simulation, benefits the research of exercise, increases the efficiency of exercise and enhances the emergency decision-making capability. This paper introduces the design and application of SAESS. (author)

Severe accident analysis code Sampson for impact project

International Nuclear Information System (INIS)

Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

2001-01-01

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

INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

D.A. Kalinich

1999-09-27

Impacts due to nearby installations and operations were determined in the Preliminary MGDS Hazards Analysis (CRWMS M&O 1996) to be potentially applicable to the proposed repository at Yucca Mountain. This determination was conservatively based on limited knowledge of the potential activities ongoing on or off the Nevada Test Site (NTS). It is intended that the Industrial/Military Activity-Initiated Accident Screening Analysis provided herein will meet the requirements of the ''Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants'' (NRC 1987) in establishing whether this external event can be screened from further consideration or must be included as a design basis event (DBE) in the development of accident scenarios for the Monitored Geologic Repository (MGR). This analysis only considers issues related to preclosure radiological safety. Issues important to waste isolation as related to impact from nearby installations will be covered in the MGR performance assessment.

Domino effect in chemical accidents: main features and accident sequences

OpenAIRE

Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

2010-01-01

The main features of domino accidents in process/storage plants and in the transportation of hazardous materials were studied through an analysis of 225 accidents involving this effect. Data on these accidents, which occurred after 1961, were taken from several sources. Aspects analyzed included the accident scenario, the type of accident, the materials involved, the causes and consequences and the most common accident sequences. The analysis showed that the most frequent causes a...

Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

International Nuclear Information System (INIS)

Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo

2012-01-01

KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described

Feedwater line break accident analysis for SMART in the view point of minimum departure from nucleate boiling ratio

Energy Technology Data Exchange (ETDEWEB)

Kim Soo Hyoung; Bae, Kyoo Hwan; Chung, Young Jong; Kim, Keung Koo [KAERI, Daejeon (Korea, Republic of)

2012-10-15

KAERI and KEPCO consortium had performed standard design of SMART(System integrated Modular Advanced ReacTor) from 2009 to 2011 and obtained standard design approval in July 2012. To confirm the safety of SMART design, all of the safety related design basis events were analyzed. A feedwater line break (FLB) is a postulated accident and is a limiting accident for a decrease in the heat removal by the secondary system in the view point of the peak RCS pressure. It is well known that departure from nucleate boiling ratio (DNBR) increases with the increase of the system pressure for conventional nuclear power plants. But SMART has comparatively lower RCS flow rate, and there is a possibility to show different DNBR behavior depending on the system pressure. To confirm that SMART is safe in case of FLB accident, the Korean nuclear regulatory body required to perform the safety analysis in the view point of minimum DNBR (MDNBR) during the licensing review process for standard design approval (SDA) of SMART design. In this paper, the safety analysis results of the FLB accident for SMART in the view point of MDNBR is described.

State-of-the-art report on accident analysis and risk analysis of reprocessing plants in European countries

International Nuclear Information System (INIS)

Nomura, Yasushi

1985-12-01

This report summarizes informations obtained from America, England, France and FRG concerning methodology, computer code, fundamental data and calculational model on accident/risk analyses of spent fuel reprocessing plants. As a result, the followings are revealed. (1) The system analysis codes developed for reactor plants can be used for reprocessing plants with some code modification. (2) Calculational models and programs have been developed for accidental phenomenological analyses in FRG, but with insufficient data to prove them. (3) The release tree analysis codes developed in FRG are available to estimate radioactivity release amount/probability via off-gas/exhaustair lines in the case of accidents. (4) The computer codes developed in America for reactor-plant environmental transport/safety analyses of released radioactivity can be applied to reprocessing facilities. (author)

Monitoring and operation system for severe accidents

International Nuclear Information System (INIS)

Fukui, Toshiki; Niida, Shinji; Kato, Yumeto

2017-01-01

Monitoring and operation system for Severe Accidents (SA-MOS) is a compact Instrumentation and Control (I and C) system developed by Mitsubishi Heavy Industries (MHI) and certificated by the Japanese Nuclear Regulatory Agency (NRA) as a design application for Japanese existing PWR nuclear power plants. The system is tailored to provide monitoring and operation for Severe Accident (SA) conditions, and consists of digitalized I and C System, Human Systems Interface (HSI) system and Power Supply (PS) system as further improvement of reliability and safety. This design plans to be applied to the next Japanese PWR plants. In accordance with the new regulatory standards that NRA has established corresponding to the Fukushima accident, a long-term Station Black Out (SBO) scenario and 24-hours power supply by the storage battery in case of SA has been required. In order to address 24-hours power supply requirement in SA condition, the storage battery volume shall be increased. However, it may be difficult to introduce additional batteries to the existing plant site because of room space constraints, etc. Therefore, power distributions for the facilities which are only used for Design Basis Accident (DBA), are shut down in order to secure 24-hours operations of facilities for SA conditions including SA-MOS. That enables efficient battery resource operations as well as optimizes room space factors shared by battery cabinets. Another benefit is to introduce dedicate HSI system for SA condition and operators shift their operations using that dedicated HSI system to cope with SA events. That can reduce operator workload which forces operators to verify or choose which controllers and indicators are available in SA conditions. Furthermore, application of SA-MOS, secures the independence of the layers (DBA⇔SA) as well as secures the plant data transfer for SA conditions outside of plant. Those plant data assets can be shared by plant operation supporting personnel and

Accident on the gas transfer system

International Nuclear Information System (INIS)

Heugel, J.

1991-10-01

An accident has happened on the Vivitron gas transfer system on the 7 th August 1991. This report presents the context, facts and inquiries, analyses the reasons and explains also how the repairing has been effected

Analysis of tritium mission FMEF/FAA fuel handling accidents

Energy Technology Data Exchange (ETDEWEB)

Van Keuren, J.C.

1997-11-18

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

CINETHICA - Core accident analysis code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

A computer program for nuclear accident analysis has been developed based on the point-kinetics approximation and one-dimensional heat transfer model for reactivity feedback calculation. Hansen's method/1/ were used for the kinetics equation solution and explicit Euler method were adopted for the thermohidraulic equations. The results were favorably compared to those from the GAPOTKIN Code/2/. (author) [pt

A route-specific system for risk assessment of radioactive materials transportation accidents

International Nuclear Information System (INIS)

Moore, J.E.; Sandquist, G.M.; Slaughter, D.M.

1995-01-01

A low-cost, powerful geographic information system (GIS) that operates on a personal computer was integrated into a software system to provide route specific assessment of the risks associated with the atmospheric release of radioactive and hazardous materials in transportation accidents. The highway transportation risk assessment (HITRA) software system described here combines a commercially available GIS (TransCAD) with appropriate models and data files for route- and accident-specific factors, such as meteorology, dispersion, demography, and health effects to permit detailed analysis of transportation risk assessment. The HITRA system allows a user to interactively select a highway or railroad route from a GIS database of major US transportation routes. A route-specific risk assessment is then performed to estimate downwind release concentrations and the resulting potential health effects imposed on the exposed population under local environmental and temporal conditions. The integration of GIS technology with current risk assessment methodology permits detailed analysis coupled with enhanced user interaction. Furthermore, HITRA provides flexibility and documentation for route planning, updating and improving the databases required for evaluating specific transportation routes, changing meteorological and environmental conditions, and local demographics

Accident analysis for US fast burst reactors

International Nuclear Information System (INIS)

Paternoster, R.; Flanders, M.; Kazi, H.

1994-01-01

In the US fast burst reactor (FBR) community there has been increasing emphasis and scrutiny on safety analysis and understanding of possible accident scenarios. This paper summarizes recent work in these areas that is going on at the different US FBR sites. At this time, all of the FBR facilities have or in the process of updating and refining their accident analyses. This effort is driven by two objectives: to obtain a more realistic scenario for emergency response procedures and contingency plans, and to determine compliance with changing regulatory standards

Probabilistic accident sequence recovery analysis

International Nuclear Information System (INIS)

Stutzke, Martin A.; Cooper, Susan E.

2004-01-01

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

President's Commission and the normal accident

International Nuclear Information System (INIS)

Perrow, C.

1982-01-01

This chapter incorporates the major points of an analysis of the accident at Three Mile Island that I prepared in September 1979. In contrast to the findings of the President's Commission (1979), I did not view the accident as the result of operator error, an inept utility, or a negligent Nuclear Regulatory Commission but as a consequence of the complexity and interdependence that characterize the system itself. I argued that the accident was inevitable-that is, that it could not have been prevented, foreseen, or quickly terminated, because it was incomprehensible. It resembled other accidents in nuclear plants and in other high risk, complex and highly interdependent operator-machine systems; none of the accidents were caused by management or operator ineptness or by poor government regulation, though these characteristics existed and should have been expected. I maintained that the accident was normal, because in complex systems there are bound to be multiple faults that cannot be avoided by planning and that operators cannot immediately comprehend

Accident analysis device for nuclear power plants

International Nuclear Information System (INIS)

Ito, Masayuki.

1982-01-01

Purpose: To enable rapid recognition of and countermeasure required for accidents upon scram, by identifying the first contact point of causes for resulting the scram and displaying the contact point of causes. Constitution: When a scram signal is inputted by way of process input device, the time of the input is determined by a timer and the contact point of causes generated just before is taken as the point whose changes occurred prior to but most closely to the generation of the signal while referring to the data memory section for the time of change of the contact point of the cause, and sent to the accident analyzing display. The accident analyzing display extracts, based on the contact point of cause, a list for the forecast accidents corresponding thereto from the data memory section and also extracts the list for the corresponding confirmation items of the accident detection and displays them together with the system from which the scram signal has been generated, the time of generation, the name of the contact point of causes operated at first, and the value of the state quantity contained in the data memory section for the store of contact point of cause at the change. (Kawakami, Y.)

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

Science.gov (United States)

Khakzad, Nima; Khan, Faisal; Amyotte, Paul

2015-07-01

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

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

International Nuclear Information System (INIS)

Sprung, J.L.; Jow, H-N; Rollstin, J.A.; Helton, J.C.

1990-12-01

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

Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

Energy Technology Data Exchange (ETDEWEB)

Saleh, J.H., E-mail: jsaleh@gatech.ed [School of Aerospace Engineering, Georgia Institute of Technology (United States); Marais, K.B. [School of Aeronautics and Astronautics, Purdue University (United States); Bakolas, E.; Cowlagi, R.V. [School of Aerospace Engineering, Georgia Institute of Technology (United States)

2010-11-15

This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations

Highlights from the literature on accident causation and system safety: Review of major ideas, recent contributions, and challenges

International Nuclear Information System (INIS)

Saleh, J.H.; Marais, K.B.; Bakolas, E.; Cowlagi, R.V.

2010-01-01

This work constitutes a short guide to the extensive but fragmented literature on accident causation and system safety. After briefly motivating the interest in accident causation and discussing the notion of a safety value chain, we delve into our multi-disciplinary review with discussions of Man Made Disasters, Normal Accident, and the High Reliability Organizations (HRO) paradigm. The HRO literature intersects an extensive literature on safety culture, a subject we then briefly touch upon. Following this discussion, we note that while these social and organizational contributions have significantly enriched our understanding of accident causation and system safety, they have important deficiencies and are lacking in their understanding of technical and design drivers of system safety and accident causation. These missing ingredients, we argue, were provided in part by the development of Probabilistic Risk Assessment (PRA). The idea of anticipating possible accident scenarios, based on the system design and configuration, as well as its technical and operational characteristics, constitutes an important contribution of PRA, which builds on and extends earlier contributions made by the development of Fault Tree and Event Tree Analysis. We follow the discussion of PRA with an exposition of the concept of safety barriers and the principle of defense-in-depth, both of which emphasize the functions and 'safety elements [that should be] deliberately inserted' along potential accident trajectories to prevent, contain, or mitigate accidents. Finally, we discuss two ideas that are emerging as foundational in the literature on system safety and accident causation, namely that system safety is a 'control problem', and that it requires a 'system theoretic' approach to be dealt with. We clarify these characterizations and indicate research opportunities to be pursued along these directions. We conclude this work with two general recommendations: (1) that more fundamental

Accident Analysis for the NIST Research Reactor Before and After Fuel Conversion

Energy Technology Data Exchange (ETDEWEB)

Baek J.; Diamond D.; Cuadra, A.; Hanson, A.L.; Cheng, L-Y.; Brown, N.R.

2012-09-30

Postulated accidents have been analyzed for the 20 MW D2O-moderated research reactor (NBSR) at the National Institute of Standards and Technology (NIST). The analysis has been carried out for the present core, which contains high enriched uranium (HEU) fuel and for a proposed equilibrium core with low enriched uranium (LEU) fuel. The analyses employ state-of-the-art calculational methods. Three-dimensional Monte Carlo neutron transport calculations were performed with the MCNPX code to determine homogenized fuel compositions in the lower and upper halves of each fuel element and to determine the resulting neutronic properties of the core. The accident analysis employed a model of the primary loop with the RELAP5 code. The model includes the primary pumps, shutdown pumps outlet valves, heat exchanger, fuel elements, and flow channels for both the six inner and twenty-four outer fuel elements. Evaluations were performed for the following accidents: (1) control rod withdrawal startup accident, (2) maximum reactivity insertion accident, (3) loss-of-flow accident resulting from loss of electrical power with an assumption of failure of shutdown cooling pumps, (4) loss-of-flow accident resulting from a primary pump seizure, and (5) loss-of-flow accident resulting from inadvertent throttling of a flow control valve. In addition, natural circulation cooling at low power operation was analyzed. The analysis shows that the conversion will not lead to significant changes in the safety analysis and the calculated minimum critical heat flux ratio and maximum clad temperature assure that there is adequate margin to fuel failure.

Accidents of loss of flow for the ETTR-2 reactor; deterministic analysis

International Nuclear Information System (INIS)

El-Messiry, A.M.

2000-01-01

The main objective for reactor safety is to keep the fuel in a thermally safe condition with adequate safety margins during all operational modes (normal-abnormal and accidental states). To achieve this purpose an accident analysis of different design base accident (DBA) as loss of flow accident (LOFA), is required assessing reactor safety. The present work concerns this transients applied to Egypt Test and Research Reactor ETRR-3 (new reactor). An accident analysis code FLOWTR is developed to investigate the thermal behaviour of the core during such flow transients. The active core is simulated by two channels: 1 - hot channel (HC), and 2 - average channel (AC) representing the remainder of the core. Each channel is divided into four axial sections. The external loop, core plenums, and core chimney are simulated by different dynamic loops. The code includes modules for pump cast down, flow regimes, decay heat, temperature distributions, and feedback coefficients. FLOWTR is verified against results from RETRAN code, THERMIC code and commissioning tests for null transient case. The comparison shows a good agreement. The study indicates that for LOFA transients, provided the scram system is available, the core is shutdown safely by low flow signal (496.6 kg/s) at 1.4 s were the HC temperature reaches the maximum value, 45.64 o C after shutdown. On the other hand, if the scram system is unavailable, and at t = 47.33 s, the core flow decreases to 67.41 kg/s, the HC temperature increases to 164.02 o C, and the HC clad surface heat flux exceeds its critical value of 400.00 W/cm 2 resulting of fuel burnout. (author)

National and regional analysis of road accidents in Spain.

Science.gov (United States)

Tolón-Becerra, A; Lastra-Bravo, X; Flores-Parra, I

2013-01-01

In Spain, the absolute fatality figures decreased almost 50 percent between 1998 and 2009. Despite this great effort, road mortality is still of great concern to political authorities. Further progress requires efficient road safety policy based on an optimal set of measures and targets that consider the initial conditions and characteristics in each region. This study attempts to analyze road accidents in Spain and its provinces in time and space during 1998-2009. First, we analyzed daily, monthly, and nationwide (NUTS 0) development of road accidents, the correlation between logarithmic transformations of road accidents and territorial and socioeconomic variables, the causality by simple linear regression of road accidents and territorial and socioeconomic variables, and preliminary frequency by fast Fourier transform. Then we analyzed the annual trend in accidents in the Spanish provinces (NUTS 3) and found a correlation between the logarithmic transformations of the mortality rate, fatalities per fatal accident, and accidents resulting in injuries per inhabitant variables and population, population density, gross domestic product (GDP), length of road network, and area. Finally, causality was analyzed by simple linear regression. The most outstanding results were the negative correlation between mortality rate and population density in Spanish provinces, which has increased over time, and that road accidents in Spain have an approximate periodicity of 57 days. The fast Fourier transform analysis of road accident frequency in Spain was useful in identifying the periodic, harmonic components of accidents and casualties. The periodicity observed both for the period 1998-2009 and by year showed that the highest intensity in road accidents was bimonthly, despite the lower number of accidents and casualties in the spectra of amplitude and power and efforts to reduce the intensity and concentration during off-season travel (summer and December).

Analysis of Child-related Road Traffic Accidents in Vietnam

Science.gov (United States)

Vu, Anh Tuan; Nguyen, Dinh Vinh Man

2018-04-01

In recent years, the number of road traffic accidents, fatalities and injuries have been decreasing, but the figures of children road traffic accidents have been increasing in Ho Chi Minh City of Vietnam. This fact strongly calls for implementing effective solutions to improve traffic safety for children by the local government. This paper presents the trends, patterns and causes of road traffic accidents involving children based on the analysis of road traffic accident data over the period 2010-2015 and the video-based observations of road traffic law violations at 15 typical school gates and 10 typical roads. The results could be useful for the city government to formulate solutions to effectively improve traffic safety for children in Ho Chi Minh City and other cities in Vietnam.

Extra-regulatory accident safety evaluation for the PWR S/F transport and storage system

International Nuclear Information System (INIS)

Seo, K. S.; Lee, J. C.; Bang, K. S.; Choi, W. S.; Lee, S. H.; Seo, J. S.; Kim, K. Y.; Jeon, J. E.

2011-06-01

In the field of high speed crash, high speed impact analyses and test were performed for two systems, the dual purpose metal cask and the concrete cask considering the aircraft crash condition. Through the tests, the procedure and methodology of the assessment were successfully validated. In the field of transient fire, the computer simulation method for transient fire was drawn through the overseas status and methodology analysis. In the field of cumulative damage evaluation for transport accident, the analysis technique for assessment for cumulative damages which occurred from successive accident conditions was developed and proposed. And the sequential tests for the dual purpose cask were performed

An analysis of the human reliability on Three Mile Island II accident considering THERP and ATHEANA methodologies

International Nuclear Information System (INIS)

Fonseca, Renato Alves da; Alvim, Antonio Carlos Marques

2005-01-01

The research on the Analysis of the Human Reliability becomes more important every day, as well as the study of the human factors and the contributions of the same ones to the incidents and accidents, mainly in complex plants or of high technology. The analysis here developed it uses the methodologies THERP (Technique for Human Error Prediction) and ATHEANA (A Technique for Human Error Analysis), as well as, the tables and the cases presented in THERP Handbook and to develop a qualitative and quantitative study of an occurred nuclear accident. The chosen accident was it of Three Mile Island (TMI). The accident analysis has revealed a series of incorrect actions that resulted in the permanent loss of the reactor and shutdown of Unit 2. This study also aims at enhancing the understanding of the THERP and ATHEANA methods and at practical applications. In addition, it is possible to understand the influence of plant operational status on human failures and the influence of human failures on equipment of a system, in this case, a nuclear power plant. (author)

The computer aided education and training system for accident management

International Nuclear Information System (INIS)

Yoneyama, Mitsuru; Kubota, Ryuji; Fujiwara, Tadashi; Sakuma, Hitoshi

1999-01-01

The education and training system for Accident Management was developed by the Japanese BWR group and Hitachi Ltd. The education and training system is composed of two systems. One is computer aided instruction (CAI) education system and the education and training system with computer simulations. Both systems are designed to be executed on personal computers. The outlines of the CAI education system and the education and training system with simulator are reported below. These systems provides plant operators and technical support center staff with the effective education and training for accident management. (author)

Fire-accident analysis code (FIRAC) verification

International Nuclear Information System (INIS)

Nichols, B.D.; Gregory, W.S.; Fenton, D.L.; Smith, P.R.

1986-01-01

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

An Analysis of Construction Accident Factors Based on Bayesian Network

OpenAIRE

Yunsheng Zhao; Jinyong Pei

2013-01-01

In this study, we have an analysis of construction accident factors based on bayesian network. Firstly, accidents cases are analyzed to build Fault Tree method, which is available to find all the factors causing the accidents, then qualitatively and quantitatively analyzes the factors with Bayesian network method, finally determines the safety management program to guide the safety operations. The results of this study show that bad condition of geological environment has the largest posterio...

System 80+TM PRA insights on severe accident prevention and mitigation

International Nuclear Information System (INIS)

Finnicum, D.J.; Jacob, M.C.; Schneider, R.E.; Weston, R.A.

2004-01-01

The System 80 + design is ABB-CE's standardized evolutionary Advanced Light Water Reactor (ALWR) design. It incorporates design enhancements based on Probabilistic Risk Assessment (PRA) insights, guidance from the ALWR Utility Requirements Document (URD), and US NRC's Severe Accident Policy. Major severe accident prevention and mitigation design features of the System 80 + design are described. The results of the System 80 + PRA are presented and the insights gained from the PRA sensitivity analyses are discussed. ABB-CE considered defense-in-depth for accident prevention and mitigation early in the design process and used robust design features to ensure that the System 80 + design achieved a low core damage frequency, low containment conditional failure probability, and excellent deterministic containment performance under severe accident conditions and to ensure that the risk was properly allocated among design features and between prevention and mitigation. (author)

Analysis of search and rescue emergency evaluation in ship accidents in Indonesia

Directory of Open Access Journals (Sweden)

Arleiny

2018-01-01

Full Text Available The objectives og this research is to describe the factors causing ship accident in Indonesia and know the effectiveness of SAR emergency in ship accident in Indonesia. The research method used in this research is qualitative research. Techniques Collection of literature study data and documents. Data validity method using triangulation. Data analysis uses interactive data analysis. The conclusions of this study are Factors that cause the occurrence of ship accidents in Indonesia, among others, the resources of the crew, the eligibility of ships, supporting facilities for shipping, operators, lack of supervision of apparatus, service users and other factors. The high number of ship accidents in Indonesia shows the ineffective implementation of SAR in ship accident in Indonesia.

Preliminary Analysis of Aircraft Loss of Control Accidents: Worst Case Precursor Combinations and Temporal Sequencing

Science.gov (United States)

Belcastro, Christine M.; Groff, Loren; Newman, Richard L.; Foster, John V.; Crider, Dennis H.; Klyde, David H.; Huston, A. McCall

2014-01-01

Aircraft loss of control (LOC) is a leading cause of fatal accidents across all transport airplane and operational classes, and can result from a wide spectrum of hazards, often occurring in combination. Technologies developed for LOC prevention and recovery must therefore be effective under a wide variety of conditions and uncertainties, including multiple hazards, and their validation must provide a means of assessing system effectiveness and coverage of these hazards. This requires the definition of a comprehensive set of LOC test scenarios based on accident and incident data as well as future risks. This paper defines a comprehensive set of accidents and incidents over a recent 15 year period, and presents preliminary analysis results to identify worst-case combinations of causal and contributing factors (i.e., accident precursors) and how they sequence in time. Such analyses can provide insight in developing effective solutions for LOC, and form the basis for developing test scenarios that can be used in evaluating them. Preliminary findings based on the results of this paper indicate that system failures or malfunctions, crew actions or inactions, vehicle impairment conditions, and vehicle upsets contributed the most to accidents and fatalities, followed by inclement weather or atmospheric disturbances and poor visibility. Follow-on research will include finalizing the analysis through a team consensus process, defining future risks, and developing a comprehensive set of test scenarios with correlation to the accidents, incidents, and future risks. Since enhanced engineering simulations are required for batch and piloted evaluations under realistic LOC precursor conditions, these test scenarios can also serve as a high-level requirement for defining the engineering simulation enhancements needed for generating them.

Future Integrated Systems Concept for Preventing Aircraft Loss-of-Control Accidents

Science.gov (United States)

Belcastro, Christine M.; Jacobson, Steven r.

2010-01-01

Loss of control remains one of the largest contributors to aircraft fatal accidents worldwide. Aircraft loss-of-control accidents are highly complex in that they can result from numerous causal and contributing factors acting alone or (more often) in combination. Hence, there is no single intervention strategy to prevent these accidents. This paper presents future system concepts and research directions for preventing aircraft loss-of-control accidents.

Analysis on Dangerous Source of Large Safety Accident in Storage Tank Area

Science.gov (United States)

Wang, Tong; Li, Ying; Xie, Tiansheng; Liu, Yu; Zhu, Xueyuan

2018-01-01

The difference between a large safety accident and a general accident is that the consequences of a large safety accident are particularly serious. To study the tank area which factors directly or indirectly lead to the occurrence of large-sized safety accidents. According to the three kinds of hazard source theory and the consequence cause analysis of the super safety accident, this paper analyzes the dangerous source of the super safety accident in the tank area from four aspects, such as energy source, large-sized safety accident reason, management missing, environmental impact Based on the analysis of three kinds of hazard sources and environmental analysis to derive the main risk factors and the AHP evaluation model is established, and after rigorous and scientific calculation, the weights of the related factors in four kinds of risk factors and each type of risk factors are obtained. The result of analytic hierarchy process shows that management reasons is the most important one, and then the environmental factors and the direct cause and Energy source. It should be noted that although the direct cause is relatively low overall importance, the direct cause of Failure of emergency measures and Failure of prevention and control facilities in greater weight.

K West integrated water treatment system subproject safety analysis document

International Nuclear Information System (INIS)

SEMMENS, L.S.

1999-01-01

This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System

K West integrated water treatment system subproject safety analysis document

Energy Technology Data Exchange (ETDEWEB)

SEMMENS, L.S.

1999-02-24

This Accident Analysis evaluates unmitigated accident scenarios, and identifies Safety Significant and Safety Class structures, systems, and components for the K West Integrated Water Treatment System.

Upgrading the electrical system of the IEA-R1 reactor to avoid triggering event of accidents

International Nuclear Information System (INIS)

Mello, Jose Roberto de; Madi Filho, Tufic

2015-01-01

The IEA-R1 research reactor at the Institute of Energy and Nuclear Research (IPEN) is a research reactor open pool type, built and designed by the American firm 'Babcox and Wilcox', having as coolant and moderator demineralized light water and Beryllium and graphite, as reflectors. The power supply system is designed to meet the electricity demand required by the loads of the reactor (Security systems and systems not related to security) in different situations the plant can meet, such as during startup, normal operation at power, shutdown, maintenance, exchange of fuel elements and accident situations. Studies have been done on possible accident initiating events and deterministic techniques were applied to assess the consequences of such incidents. Thus, the methods used to identify and select the accident initiating events, the methods of analysis of accidents, including sequence of events, transient analysis and radiological consequences, have been described. Finally, acceptance criteria of radiological doses are described. Only a brief summary of the item concerning loss of electrical power will be presented. The loss of normal electrical power at the IEA-R1 reactor is very common. In the case of Electric External Power Loss, at the IEA-R1 reactor building, there may be different sequences of events, as described below. When the supply of external energy in the IEA-R1 facility fails, the Electrical Distribution Vital System, consisting of 4 (four) generators type 'UPS', starts operation, immediately and it will continue supplying power to the reactor control table, core cooling system and other security systems. To contribute to security, in the electric power failure, starts to operate the Emergency Cooling System (SRE). SRE has the function of removing residual heat from the core to prevent the melting of fuel elements in the event of loss of refrigerant to the core. Adding to the generators with batteries group system, new auxiliary

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

Lee, Yongjae; Seo, Seungwon; Kim, Sung Joong; Ha, Kwang Soon; Kim, Hwan-Yeol

2014-01-01

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

A methodology for accident analysis of fusion breeder blankets and its application to helium-cooled lead–lithium blanket

International Nuclear Information System (INIS)

Panayotov, Dobromir; Poitevin, Yves; Grief, Andrew; Trow, Martin; Dillistone, Michael

2016-01-01

'Fusion for Energy' (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. Furthermore, the methodology phases are illustrated in the paper by its application to the EU HCLL TBS using both MELCOR and RELAP5 codes.

Response of Soviet VVER-440 accident localization systems to overpressurization

International Nuclear Information System (INIS)

Kulak, R.F.; Fiala, C.; Sienicki, J.J.

1989-01-01

The Soviet designed VVER-440 model V230 and VVER-440 model V213 reactors do not use full containments to mitigate the effects of accidents. Instead, these VVER-440 units employ a sealed set of interconnected compartments, collectively called the accident localization system (ALS), to reduce the release of radionuclides to the atmosphere during accidents. Descriptions of the VVER accident localization structures may be found in the report DOE NE-0084. The objective of this paper is to evaluate the structural integrity of the VVER-440 ALS at the Soviet design pressure, and to determine their response to pressure loadings beyond the design value. Complex, three-dimensional, nonlinear, finite element models were developed to represent the major structural components of the localization systems of the VVER-440 models V230 and V213. The interior boundary of the localization system was incrementally pressurized in the calculations until the prediction of gross failure. 6 refs., 9 figs

Accident rate analysis for well drilling at Kuban' morneftegazprom association

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, V B; Kezchikov, A V

1981-01-01

Analysis of emergency procedures at the association during 1976--1977 is provided. Conclusions were made and plans established for basic directions of engineering-production operations and for association and enterprise division sections with regard to lowering accident rate and time period necessary to eliminate accidents.

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

Energy Technology Data Exchange (ETDEWEB)

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

1991-12-15

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

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

International Nuclear Information System (INIS)

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

1991-12-01

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

Comparative analysis of station blackout accident progression in typical PWR, BWR, and PHWR

International Nuclear Information System (INIS)

Park, Soo Young; Ahn, Kwang Il

2012-01-01

Since the crisis at the Fukushima plants, severe accident progression during a station blackout accident in nuclear power plants is recognized as a very important area for accident management and emergency planning. The purpose of this study is to investigate the comparative characteristics of anticipated severe accident progression among the three typical types of nuclear reactors. A station blackout scenario, where all off-site power is lost and the diesel generators fail, is simulated as an initiating event of a severe accident sequence. In this study a comparative analysis was performed for typical pressurized water reactor (PWR), boiling water reactor (BWR), and pressurized heavy water reactor (PHWR). The study includes the summarization of design differences that would impact severe accident progressions, thermal hydraulic/severe accident phenomenological analysis during a station blackout initiated-severe accident; and an investigation of the core damage process, both within the reactor vessel before it fails and in the containment afterwards, and the resultant impact on the containment.

Insights Gained from Forensic Analysis with MELCOR of the Fukushima-Daiichi Accidents.

Energy Technology Data Exchange (ETDEWEB)

Andrews, Nathan C. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Gauntt, Randall O. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2017-10-01

Since the accidents at Fukushima-Daiichi, Sandia National Laboratories has been modeling these accident scenarios using the severe accident analysis code, MELCOR. MELCOR is a widely used computer code developed at Sandia National Laboratories since ~1982 for the U.S. Nuclear Regulatory Commission. Insights from the modeling of these accidents is being used to better inform future code development and potentially improved accident management. To date, our necessity to better capture in-vessel thermal-hydraulic and ex-vessel melt coolability and concrete interactions has led to the implementation of new models. The most recent analyses, presented in this paper, have been in support of the of the Organization for Economic Cooperation and Development Nuclear Energy Agency’s (OECD/NEA) Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project. The goal of this project is to accurately capture the source term from all three releases and then model the atmospheric dispersion. In order to do this, a forensic approach is being used in which available plant data and release timings is being used to inform the modeled MELCOR accident scenario. For example, containment failures, core slumping events and lower head failure timings are all enforced parameters in these analyses. This approach is fundamentally different from a blind code assessment analysis often used in standard problem exercises. The timings of these events are informed by representative spikes or decreases in plant data. The combination of improvements to the MELCOR source code resulting from analysis previous accident analysis and this forensic approach has allowed Sandia to generate representative and plausible source terms for all three accidents at Fukushima Daiichi out to three weeks after the accident to capture both early and late releases. In particular, using the source terms developed by MELCOR, the MACCS software code, which models atmospheric dispersion and

A systems approach to the management of radiation accidents

International Nuclear Information System (INIS)

Richter, L.L.; Berk, H.W.; Teates, C.D.; Larkham, N.E.; Friesen, E.J.; Edlich, R.F.

1980-01-01

Management of radiation accident patients should have a multidisciplinary approach that includes all health professionals as well as members of public safety agencies. Emergency plans for radiation accidents include detection of the ionizing radiation, patient evacuation, resuscitation, and decontamination. The resuscitated patient should be transported to a radiation control area located outside but adjacent to the emergency department. Ideally this area is accessed through an entrance separate from that used for the main flow of daily emergency department patients. The hospital staff, provided with protective clothing, dosimeters, and preprinted guidelines, continues the resuscitation and definitive care of the patient. This system approach to the management of radiation accidents may be tailored to meet the specific needs of other emergency medical systems

Fuel temperature analysis method for channel-blockage accident in HTTR

International Nuclear Information System (INIS)

Maruyama, So; Fujimoto, Nozomu; Sudo, Yukio; Kiso, Yoshihiro; Hayakawa, Hitoshi

1994-01-01

During operation of the High Temperature Engineering Test Reactor (HTTR), coolability must be maintained without core damage under all postulated accident conditions. Channel blockage of a fuel element was selected as one of the design-basis accidents in the safety evaluation of the reactor. The maximum fuel temperature for such a scenario has been evaluated in the safety analysis and is compared to the core damage limits.For the design of the HTTR, an in-core thermal and hydraulic analysis code ppercase[flownet/trump] was developed. This code calculates fuel temperature distribution, not only for a channel blockage accident but also for transient conditions. The validation of ppercase[flownet/trump] code was made by comparison of the analytical results with the results of thermal and hydraulic tests by the Helium Engineering Demonstration Loop (HENDEL) multi-channel test rig (T 1-M ), which simulated one fuel column in the core. The analytical results agreed well with the experiments in which the HTTR operating conditions were simulated.The maximum fuel temperature during a channel blockage accident is 1653 C. Therefore, it is confirmed that the integrity of the core is maintained during a channel blockage accident. ((orig.))

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

Energy Technology Data Exchange (ETDEWEB)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident.

Primary pipe rupture accident analysis for the Clinch River Breeder Reactor

International Nuclear Information System (INIS)

Albright, D.C.; Bari, R.A.

1976-04-01

In this report, the thermal transient response of the CRBR to a severe primary coolant flow perturbation, initiated by a rupture of the primary heat transport system piping, is analyzed. This hypothetical accident is studied under the further assumption that the plant protection system does function according to current design descriptions for the CRBR. Although a brief discussion of an unprotected (no scram) pipe rupture accident is presented, the major emphasis of the present report is on the protected accident

Comparison of Management Oversight and Risk Tree and Tripod-Beta in Excavation Accident Analysis

Directory of Open Access Journals (Sweden)

Mohamadfam

2015-01-01

Full Text Available Background Accident investigation programs are a necessary part in identification of risks and management of the business process. Objectives One of the most important features of such programs is the analysis technique for identifying the root causes of accidents in order to prevent their recurrences. Analytical Hierarchy Process (AHP was used to compare management oversight and risk tree (MORT with Tripod-Beta in order to determine the superior technique for analysis of fatal excavation accidents in construction industries. Materials and Methods MORT and Tripod-Beta techniques were used for analyzing two major accidents with three main steps. First, these techniques were applied to find out the causal factors of the accidents. Second, a number of criteria were developed for the comparison of the techniques and third, using AHP, the techniques were prioritized in terms of the criteria for choosing the superior one. Results The Tripod-Beta investigation showed 41 preconditions and 81 latent causes involved in the accidents. Additionally, 27 root causes of accidents were identified by the MORT analysis. Analytical hierarchy process (AHP investigation revealed that MORT had higher priorities only in two criteria than Tripod-Beta. Conclusions Our findings indicate that Tripod-Beta with a total priority of 0.664 is superior to MORT with the total priority of 0.33. It is recommended for future research to compare the available accident analysis techniques based on proper criteria to select the best for accident analysis.

Characteristics of Hydrogen Monitoring Systems for Severe Accident Management at a Nuclear Power Plant

Science.gov (United States)

Petrosyan, V. G.; Yeghoyan, E. A.; Grigoryan, A. D.; Petrosyan, A. P.; Movsisyan, M. R.

2018-02-01

One of the main objectives of severe accident management at a nuclear power plant is to protect the integrity of the containment, for which the most serious threat is possible ignition of the generated hydrogen. There should be a monitoring system providing information support of NPP personnel, ensuring data on the current state of a containment gaseous environment and trends in its composition changes. Monitoring systems' requisite characteristics definition issues are considered by the example of a particular power unit. Major characteristics important for proper information support are discussed. Some features of progression of severe accident scenarios at considered power unit are described and a possible influence of the hydrogen concentration monitoring system performance on the information support reliability in a severe accident is analyzed. The analysis results show that the following technical characteristics of the combustible gas monitoring systems are important for the proper information support of NPP personnel in the event of a severe accident at a nuclear power plant: measured parameters, measuring ranges and errors, update rate, minimum detectable concentration of combustible gas, monitoring reference points, environmental qualification parameters of the system components. For NPP power units with WWER-440/270 (230) type reactors, which have a relatively small containment volume, the update period for measurement results is a critical characteristic of the containment combustible gas monitoring system, and the choice of monitoring reference points should be focused not so much on the definition of places of possible hydrogen pockets but rather on the definition of places of a possible combustible mixture formation. It may be necessary for the above-mentioned power units to include in the emergency operating procedures measures aimed at a timely heat removal reduction from the containment environment if there are signs of a severe accident phase

A modular structure to accident identification using neural networks

International Nuclear Information System (INIS)

Duque Estrada, Cassius Rodrigo

2005-01-01

This work uses the accident identification method based on Artificial Neural Networks (ANN) as basic blocks of a modular structure, allowing the inclusion of new accidents to be identified without modifying the ANN already trained. This structure comprises several modules for accident identification and one module for analysis. Each identification module follows the structure of the basic block. The identification modules are responsible for the recognition of an accident belonging to the specific set of events for which it were trained. The analysis module processes the output from the identification module to determine the system response. In order to test this structure it was proposed a transient identification problem comprising fifty accidents distributed in five identification modules. The results have demonstrated that the accident identification method used as basic block of a modular structure allows the inclusion of new sets of accidents, or variations of a same accident, without modifying the ANN already trained. For this, it is enough to include into the system an specific module for this new set of accidents. (author)

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Accident and emergency management

International Nuclear Information System (INIS)

Andersen, V.; Moellenbach, K.; Heinonen, R.; Jakobsson, S.; Kukko, T.; Berg, Oe.; Larsen, J.S.; Westgaard, T.; Magnusson, B.; Andersson, H.; Holmstroem, C.; Brehmer, B.; Allard, R.

1988-06-01

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

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

Science.gov (United States)

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

2014-12-01

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

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

The Swedish regulatory approach to safety assessment and severe accident management is briefly described. The safety assessment program, which focuses on prevention of incidents and accidents, has three main components: periodic safety reviews, probabilistic safety analysis, and analysis of postulated disturbances and accident progression sequences. Management and man-technology-organisation issues, as well as inspections, play a key role in safety assessment. Basis for severe accident management were established by the Government decisions in 1981 and 1986. By the end of 1988, the severe accident mitigation systems and emergency operating procedures were implemented at all Swedish reactors. The severe accident research has continued after 1988 for further verification of the protection provided by the systems and reduction of remaining uncertainties in risk dominant phenomena

Detection and analysis of accident black spots with even small accident figures.

NARCIS (Netherlands)

Oppe, S.

1982-01-01

Accident black spots are usually defined as road locations with high accident potentials. In order to detect such hazardous locations we have to know the probability of an accident for a traffic situation of some kind, or the mean number of accidents for some unit of time. In almost all procedures

Hanford Waste Tank Bump Accident and Consequence Analysis

International Nuclear Information System (INIS)

BRATZEL, D.R.

2000-01-01

This report provides a new evaluation of the Hanford tank bump accident analysis and consequences for incorporation into the Authorization Basis. The analysis scope is for the safe storage of waste in its current configuration in single-shell and double-shell tanks

Loss-of-Coolant and Loss-of-Flow Accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1995-01-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Loss-of-coolant and loss-of-flow accidents in the SEAFP first wall/blanket cooling system

International Nuclear Information System (INIS)

Komen, E.M.J.; Koning, H.

1994-07-01

This paper presents the RELAP5/MOD3 thermal-hydraulic analysis of three Loss-of-Coolant Accidents (LOCAs) and three Loss-of-Flow Accidents (LOFAs) in the first wall/blanket cooling system of the SEAFP reactor design. The analyses deal with the transient thermal-hydraulic behaviour inside the cooling systems and the temperature development inside the nuclear components. As it appears, the temperature increase in the first wall Be-coating is limited to 30 K when an emergency plasma shutdown is initiated within 10 s following pump trip. (orig.)

Cause-effect analysis on Fukushima accident reports. What did McMaster undergraduate students learn?

International Nuclear Information System (INIS)

Nagasaki, Shinya

2016-01-01

In the ENG PHYS 4ES3 Course “Special Topics in Energy Systems (2014-2015)” in McMaster University, sixteen 4th-year undergraduate students studied the Fukushima accident, discussed the causes of accident and its impacts on the energy systems from the sustainability point of view, made the oral presentation and submitted the reports. In this paper, a cause-effect and causal-loop analysis was applied to the discussion in the reports, the diagram of cause-effect relationship was drawn, and the important problems were extracted from the diagram. It was found that the important problems and the diagram of cause-effect relationship McMaster undergraduate students considered were similar to the essential problems and the diagram Horii pointed out, although Interim Report of the Investigation Committee on the Accident at Fukushima Nuclear Stations of Tokyo Electric Power Company which Horii used was not adopted in the reports submitted by students. (author)

Applicability of simplified human reliability analysis methods for severe accidents

Energy Technology Data Exchange (ETDEWEB)

Boring, R.; St Germain, S. [Idaho National Lab., Idaho Falls, Idaho (United States); Banaseanu, G.; Chatri, H.; Akl, Y. [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada)

2016-03-15

Most contemporary human reliability analysis (HRA) methods were created to analyse design-basis accidents at nuclear power plants. As part of a comprehensive expansion of risk assessments at many plants internationally, HRAs will begin considering severe accident scenarios. Severe accidents, while extremely rare, constitute high consequence events that significantly challenge successful operations and recovery. Challenges during severe accidents include degraded and hazardous operating conditions at the plant, the shift in control from the main control room to the technical support center, the unavailability of plant instrumentation, and the need to use different types of operating procedures. Such shifts in operations may also test key assumptions in existing HRA methods. This paper discusses key differences between design basis and severe accidents, reviews efforts to date to create customized HRA methods suitable for severe accidents, and recommends practices for adapting existing HRA methods that are already being used for HRAs at the plants. (author)

Exploring the potential of data mining techniques for the analysis of accident patterns

DEFF Research Database (Denmark)

Prato, Carlo Giacomo; Bekhor, Shlomo; Galtzur, Ayelet

2010-01-01

Research in road safety faces major challenges: individuation of the most significant determinants of traffic accidents, recognition of the most recurrent accident patterns, and allocation of resources necessary to address the most relevant issues. This paper intends to comprehend which data mining...... and association rules) data mining techniques are implemented for the analysis of traffic accidents occurred in Israel between 2001 and 2004. Results show that descriptive techniques are useful to classify the large amount of analyzed accidents, even though introduce problems with respect to the clear...... importance of input and intermediate neurons, and the relative importance of hundreds of association rules. Further research should investigate whether limiting the analysis to fatal accidents would simplify the task of data mining techniques in recognizing accident patterns without the “noise” probably...

Realistic analysis of steam generator tube rupture accident in Angra-1 reactor

International Nuclear Information System (INIS)

Fontes, S.W.F.

1989-01-01

This paper presents the analysis of different scenarios for a Steam Generator Tube Rupture accident (SGTR) in Angra-1 NPP. The results and conclusions will be used as support in the preparation of the emergency situation programs for the plant. For the analysis a SGTR simulation was performed with RETRAN-02 code. The results indicated that the core integrity and the plant itself will not affect by small ruptures in SG tubes. For large ruptures the analysis demonstrated that the accident may have harmful consequences if the operator do not actuate effectively since the initial moments of the accidents. (author) [pt

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

International Nuclear Information System (INIS)

Wang Xingyu; Shi Zhongqi

2002-01-01

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

System thermalhydraulics for design basis accident analysis and simulation: Status of tools and methods and direction for future R&D

Energy Technology Data Exchange (ETDEWEB)

Bestion, D., E-mail: dominique.bestion@cea.fr

2017-02-15

Highlights: • A state of the art on system code application is presented. • Requirements for demonstration of code up-scaling capabilities are proposed. • The role of multi-scale analysis in safety analysis is explained. • Uncertainty quantifications methodologies for system codes and CFD codes are compared and discussed. - Abstract: System thermalhydraulic investigations of Design Basis Accident require several tools and methods including the Process Identification and Ranking Table, the scaling, experiment analysis, modelling, code development, code Validation and Verification, and Uncertainty Quantification. This paper intends to give an overview of these methods and tools showing what the state of the art is, and presenting some recent advances. Recommendations are made with future direction for R&D including the need of new advanced experiments and instrumentation, and the future role of CFD and multi-scale analyses. For many people it is not clear what current system codes are, and what they can be. Then the main characteristics of these codes are recalled and propositions are made to clarify the code capabilities and limitations and to improve the knowledge of the conditions for a correct application of the codes for safety in a Best Estimate Plus Uncertainty approach. Also, the on-going developments of 3-field models and Transport of Interfacial Area are summarized and associated experimental needs are identified. The growing role of 3D modelling of reactor core and Pressure Vessel requires additional experimental data for a proper validation. CFD in open medium also contributes to investigations when 3D geometrical aspects play an important role. Recent activities performed in the OECD-NEA Working Group for Analysis and Management of Accidents is summarized and recent applications of two-phase CFD to boiling flows and two-phase PTS scenarios are reported. The role of a multi-scale approach for safety issues is illustrated with the LOCA transients

Nuclear-station post-accident liquid-sampling system: developed by Duke Power Company

International Nuclear Information System (INIS)

Burton, D.A.; Birch, M.L.; Orth, W.C.

1981-01-01

The accident at Three Mile Island showed that means must be provided to determine the radioactivity levels in high activity liquid and gaseous systems of a nuclear power plant without undue radiation exposure to personnel. The Duke Power Post Accident Liquid Sampling System provides the means for obtaining diluted liquid samples and diluted dissolved gas samples following a reactor accident involving substantial core damage. Their approach yields a straightforward engineering solution at a fraction of the cost of other systems. A description of the system, general design criteria, and color coded flow diagrams are included

Impact Analysis for Risks in Informatics Systems

OpenAIRE

Baicu, Floarea; Baches, Maria Alexandra

2013-01-01

In this paper are presented methods of impact analysis on informatics system security accidents, qualitative and quantitative methods, starting with risk and informational system security definitions. It is presented the relationship between the risks of exploiting vulnerabilities of security system, security level of these informatics systems, probability of exploiting the weak points subject to financial losses of a company, respectively impact of a security accident on the company. Herewit...

Human factors analysis of incident/accident report

International Nuclear Information System (INIS)

Kuroda, Isao

1992-01-01

Human factors analysis of accident/incident has different kinds of difficulties in not only technical, but also psychosocial background. This report introduces some experiments of 'Variation diagram method' which is able to extend to operational and managemental factors. (author)

Analysis of National Major Work Safety Accidents in China, 2003-2012.

Science.gov (United States)

Ye, Yunfeng; Zhang, Siheng; Rao, Jiaming; Wang, Haiqing; Li, Yang; Wang, Shengyong; Dong, Xiaomei

2016-01-01

This study provides a national profile of major work safety accidents in China, which cause more than 10 fatalities per accident, intended to provide scientific basis for prevention measures and strategies to reduce major work safety accidents and deaths. Data from 2003-2012 Census of major work safety accidents were collected from State Administration of Work Safety System (SAWS). Published literature and statistical yearbook were also included to implement information. We analyzed the frequency of accidents and deaths, trend, geographic distribution and injury types. Additionally, we discussed the severity and urgency of emergency rescue by types of accidents. A total of 877 major work safety accidents were reported, resulting in 16,795 deaths and 9,183 injuries. The numbers of accidents and deaths, mortality rate and incidence of major accidents have declined in recent years. The mortality rate and incidence was 0.71 and 1.20 per 10(6) populations in 2012, respectively. Transportation and mining contributed to the highest number of major accidents and deaths. Major aviation and railway accidents caused more casualties per incident, while collapse, machinery, electrical shock accidents and tailing dam accidents were the most severe situation that resulted in bigger proportion of death. Ten years' major work safety accident data indicate that the frequency of accidents and number of eaths was declined and several safety concerns persist in some segments.

MELCOR Severe Accident Analysis on the SMART Reactor

International Nuclear Information System (INIS)

Kim, Tae Woon; Jin, Young Ho; Kim, Young In; Kim, Keung Koo; Wang, Ziao; Revankar, Shripad

2014-01-01

A severe accident is analyzed for Korea SMR reactor, SMART. Core melt down sequences are analyzed for SMART reactor core using MELCOR version 1.8.5. MELCOR is developed by Sandia National Laboratory for US NRC for the simulation of severe accidents in nuclear power plants. Two cases are simulated here and compared between them; one is the case for core having 3 concentric rings and the other is the case for core having 5 concentric rings. One inch break LOCA scenario is simulated and compared between these two core models. Time sequences for the thermal hydraulic behaviors of RPV and thermal heatup behaviors of reactor core are explained in graphically. Thermal hydraulic behavior such as the change of pressure, level, mass, and temperature of RPV is explained. Thermal heatup behavior of reactor core such as oxidation of cladding, hydrogen generation, core slumping down to lower plenum, and finally creep rupture of PRV lower head is explained. Engineered safety features such as safety injection systems (SIS), and Passive residual heat removal systems (PHRS), etc. are assumed to be not working. One inch break of severe accident is simulated on Korean SMR (SMART) Integral PWR with MELCOR code version 1.8.5. Core melt progression and lower head failure time is very slow compared to other commercial reactors. Simulation on 3 and 5 radial rings core models gives very similar pattern in core cell failure timings. Other various accident scenarios (for example, SBO in Fukushima) will be tried further. Containment behaviors and source term behaviors in severe accident conditions will be analyzed in future

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

Kobayashi, Masahide; Narabayashi, Tadashi; Tsuji, Masashi; Chiba, Go; Nagata, Yasunori; Shimoe, Tomohiro

2015-01-01

As a result of the Great East Japan Earthquake that occurred on 11 March 2011, all AC and DC power at the Fukushima Daiichi NPP units 1 to 3 were lost soon after the tsunami. The core cooling function was lost, and the cores of units 1 to 3 were damaged. The purpose of this work is to clarify the progress of the accident in unit 1, which was damaged the earliest among the 3 units. Therefore, an original severe accident analysis code was developed, and the progress of the accident was evaluated from the analysis results and the actual data. As a result, the leakage path from a pressure vessel was clarified, and some lessons and knowledge were gained. (author)

Analysis of system and of course of events

International Nuclear Information System (INIS)

Hoertner, H.; Kersting, E.J.; Puetter, B.M.

1986-01-01

The analysis of the system and of the course of events is used to determine the frequency of core melt-out accidents and to describe the safety-related boundary conditions of appropriate accidents. The lecture is concerned with the effect of system changes in the reference plant and the effect of triggering events not assessed in detail or not sufficiently assessed in detail in phase A of the German Risk Study on the frequency of core melt-out accidents, the minimum requirements for system functions for controlling triggering events, i.e. to prevent core melt-out accidents, the reliability data important for reliability investigations and frequency assessments. (orig./DG) [de

Analysis of the effects of the pressure wave generated in loss of coolant accidents in reactor vessels

International Nuclear Information System (INIS)

Valero Martinez, M.

1980-01-01

The increasing demands in the field of ''Nuclear Safety'', obliges to a perfect knowledge of the causes and effects of every possible accident in a nuclear power plant. In this paper will be analysed the effects of the pressure wave appearing in a LOCA (Loss of collant accident). The pressure wave could deform the following structures: core barrel wall, cover and bottom, control rods and safety coolant system. Any change of the geometry of these structures could provoke and incorrect system reaction after the accident has happened. The basis and hypothesis for the theoretical analysis will be exposed. The structures are considered to be rigid. A typical boiling water be analysed and the developed theory will be verified in comparations with experimental results and the results obtained with some others models. Due to the easy application and short calculation time of the created programmes, they are recommended for parametrical calculations in the analysis of the pressurized water reactors and boiling water reactors. (author)

Circuit board accident--organizational dimension hidden by prescribed safety.

Science.gov (United States)

de Almeida, Ildeberto Muniz; Buoso, Eduardo; do Amaral Dias, Maria Dionísia; Vilela, Rodolfo Andrade Gouveia

2012-01-01

This study analyzes an accident in which two maintenance workers suffered severe burns while replacing a circuit breaker panel in a steel mill, following model of analysis and prevention of accidents (MAPA) developed with the objective of enlarging the perimeter of interventions and contributing to deconstruction of blame attribution practices. The study was based on materials produced by a health service team in an in-depth analysis of the accident. The analysis shows that decisions related to system modernization were taken without considering their implications in maintenance scheduling and creating conflicts of priorities and of interests between production and safety; and also reveals that the lack of a systemic perspective in safety management was its principal failure. To explain the accident as merely non-fulfillment of idealized formal safety rules feeds practices of blame attribution supported by alibi norms and inhibits possible prevention. In contrast, accident analyses undertaken in worker health surveillance services show potential to reveal origins of these events incubated in the history of the system ignored in practices guided by the traditional paradigm.

Process hazards analysis (PrHA) program, bridging accident analyses and operational safety

International Nuclear Information System (INIS)

Richardson, J.A.; McKernan, S.A.; Vigil, M.J.

2003-01-01

Recently the Final Safety Analysis Report (FSAR) for the Plutonium Facility at Los Alamos National Laboratory, Technical Area 55 (TA-55) was revised and submitted to the US. Department of Energy (DOE). As a part of this effort, over seventy Process Hazards Analyses (PrHAs) were written and/or revised over the six years prior to the FSAR revision. TA-55 is a research, development, and production nuclear facility that primarily supports US. defense and space programs. Nuclear fuels and material research; material recovery, refining and analyses; and the casting, machining and fabrication of plutonium components are some of the activities conducted at TA-35. These operations involve a wide variety of industrial, chemical and nuclear hazards. Operational personnel along with safety analysts work as a team to prepare the PrHA. PrHAs describe the process; identi fy the hazards; and analyze hazards including determining hazard scenarios, their likelihood, and consequences. In addition, the interaction of the process to facility systems, structures and operational specific protective features are part of the PrHA. This information is rolled-up to determine bounding accidents and mitigating systems and structures. Further detailed accident analysis is performed for the bounding accidents and included in the FSAR. The FSAR is part of the Documented Safety Analysis (DSA) that defines the safety envelope for all facility operations in order to protect the worker, the public, and the environment. The DSA is in compliance with the US. Code of Federal Regulations, 10 CFR 830, Nuclear Safety Management and is approved by DOE. The DSA sets forth the bounding conditions necessary for the safe operation for the facility and is essentially a 'license to operate.' Safely of day-to-day operations is based on Hazard Control Plans (HCPs). Hazards are initially identified in the PrI-IA for the specific operation and act as input to the HCP. Specific protective features important to worker

Problem of corium melt coolability in passive protection systems against severe accidents in the containment

Directory of Open Access Journals (Sweden)

Ali Kalvand

2018-05-01

Full Text Available Paper is devoted to the development of the mathematical model and analysis of the problem of corium melt interaction with low-temperature melting blocks in the passive protection systems against severe accidents at the NPP, which is of high importance for substantiation of the nuclear power safety, for building and successful op-erating of passive protection systems. In the third-generation reactors passive protection systems against severe accidents at the NPP are mandatory, therefore this paper is of importance for the nuclear power safety. A few configurations for the cooling blocks’ distribution have been considered and an analysis of the blocks’ melting and corium’s cooling in the pool under reactor vessel have been done, which can serve more effective for further improvement of the safety current systems and for the development of new ones. The ways for solution of the problems and the methods for their successful elaboration were discussed. The developed mathematical models and the analysis performed in the paper might be helpful for the design of passive protection systems of the cori-um melt retention inside the containment after corium melt eruption from the broken reactor vessel.

Analyses of severe accident scenarios in RBMK-1500

International Nuclear Information System (INIS)

Kaliatka, A.; Rimkevicius, S.; Uspuras, E.; Urbonavicius, E.

2006-01-01

Even though research of severe accidents in light water reactors is performed around the world for several decades many questions remain. Research is mostly performed for vessel-type reactors. RBMK is a channel type light water reactor, which differs from the vessel-type reactors in several aspects. These differences impose some specifics in the accident phenomena and processes that occur during severe accidents. Severe accident research for RBMK reactors is taking first steps and very little information is available in the open literature. The existing severe accident analysis codes are developed for vessel-type reactors and their application to the analysis of accidents in RBMK is not straightforward. This paper presents the results of an analysis of large loss-of-coolant accident scenarios with loss of coolant injection to the core of RBMK-1500. The analysis performed considers processes in the reactor core, in the reactor cooling system and in the confinement until the fuel melting started. This paper does not aim to answer all the questions regarding severe accidents in RBMK but rather to start a discussion, identify the expected timing of the key phenomena. (orig.)

Human factors review for nuclear power plant severe accident sequence analysis

International Nuclear Information System (INIS)

Krois, P.A.; Haas, P.M.

1985-01-01

The paper discusses work conducted to: (1) support the severe accident sequence analysis of a nuclear power plant transient based on an assessment of operator actions, and (2) develop a descriptive model of operator severe accident management. Operator actions during the transient are assessed using qualitative and quantitative methods. A function-oriented accident management model provides a structure for developing technical operator guidance on mitigating core damage preventing radiological release

Use of inelastic analysis to determine the response of packages to puncture accidents

International Nuclear Information System (INIS)

Ammerman, D.J.; Ludwigsen, J.S.

1996-01-01

The accurate analytical determination of the response of radioactive material transportation packages to the hypothetical puncture accident requires inelastic analysis techniques. Use of this improved analysis method recudes the reliance on empirical and approximate methods to determine the safety for puncture accidents. This paper will discuss how inelastic analysis techniques can be used to determine the stresses, strains and deformations resulting from puncture accidents for thin skin materials with different backing materials. A method will be discussed to assure safety for all of these types of packages

Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

International Nuclear Information System (INIS)

Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu

2015-01-01

Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance

Consequence Analysis of Release from KN-18 Cask during a Severe Transportation Accident

Energy Technology Data Exchange (ETDEWEB)

Lim, Heoksoon; Bhang, Giin; Na, Janghwan; Ban, Jaeha; Kim, Myungsu [KAERI, Daejeon (Korea, Republic of)

2015-05-15

Korea Hydro and Nuclear Power (KHNP) has launched a project entitled 'Development of APR1400 Physical Protection System Design' and conducting a new conceptual physical protection system(PPS) design. One of mayor contents is consequence analysis for spent nuclear fuel cask. Proper design of physical protection system for facilities and storage and transformation involving nuclear and radioactive material requires the quantification of potential consequence from prescribed sabotage and theft scenarios in order to properly understand the level of PPS needed for specific facilities and materials. An important aspect of the regulation of the nuclear industry is assessing the risk to the public and the environment from a release of radioactive material produced by accidental or intentional scenarios. This paper describes the consequence analysis methodology, structural analysis for KN-18 cask and results of release from the cask during a severe transportation accident. Accident during spent fuel cask transportation was numerically calculated for KN-18, and showed the integrity of the fuel assemblies and cask itself was unharmed on a scenario that is comparable to state of art NRC research. Even assumption of leakage as a size of 1 x 10''2 mm''2 does not exceed for a certain criteria at any distance.