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.

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.

Safety analysis of accident localization system

International Nuclear Information System (INIS)

1999-01-01

A complex safety analysis of accident localization system of Ignalina NPP was performed. Calculation results obtained, results of non-destruct ing testing and experimental data of reinforced concrete testing of buildings does not revealed deficiencies of buildings of accident localization system at unit 1 of Ignalina NPP. Calculations were performed using codes NEPTUNE, ALGOR, CONTAIN

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

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)

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)

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...

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

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

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

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

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

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.

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.

Limitations of systemic accident analysis methods

Directory of Open Access Journals (Sweden)

Casandra Venera BALAN

2016-12-01

Full Text Available In terms of system theory, the description of complex accidents is not limited to the analysis of the sequence of events / individual conditions, but highlights nonlinear functional characteristics and frames human or technical performance in relation to normal functioning of the system, in safety conditions. Thus, the research of the system entities as a whole is no longer an abstraction of a concrete situation, but an exceeding of the theoretical limits set by analysis based on linear methods. Despite the issues outlined above, the hypothesis that there isn’t a complete method for accident analysis is supported by the nonlinearity of the considered function or restrictions, imposing a broad vision of the elements introduced in the analysis, so it can identify elements corresponding to nominal parameters or trigger factors.

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.

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.

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

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

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

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

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.

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.

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.

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.

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).

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

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.

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.

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)

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 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.)

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.

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

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 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.

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

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)

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.

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

Methods for air cleaning system design and accident analysis

International Nuclear Information System (INIS)

Gregory, W.S.; Nichols, B.D.

1987-01-01

This paper describes methods, in the form of a handbook and five computer codes, that can be used for nuclear facility air cleaning system design and accident analysis. Four of the codes were developed primarily at the Los Alamos National Laboratory, and one was developed in France. Tools such as these are used to design ventilation systems in the mining industry but do not seem to be commonly used in the nuclear industry. For example, the Nuclear Air Cleaning Handbook is an excellent design reference, but it fails to include information on computer codes that can be used to aid in the design process. These computer codes allow the analyst to use the handbook information to form all the elements of a complete system design. Because these analysis methods are in the form of computer codes they allow the analyst to investigate many alternative designs. In addition, the effects of many accident scenarios on the operation of the air cleaning system can be evaluated. These tools originally were intended for accident analysis, but they have been used mostly as design tools by several architect-engineering firms. The Cray, VAX, and personal computer versions of the codes, an accident analysis handbook, and the codes availability will be discussed. The application of these codes to several design operations of nuclear facilities will be illustrated, and their use to analyze the effect of several accident scenarios also will be described

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

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.

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.

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.

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

[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

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.

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)

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...

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)

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

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

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

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)

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

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.

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)

A System Supporting the Analysis of Motorway Traffic Accidents

Directory of Open Access Journals (Sweden)

Davide Anghinolfi

2015-12-01

Full Text Available This work presents a business intelligence tool for monitoring traffic accidents on motorways and supporting decisions relevant to road safety. The system manages information on road characteristics, traffic accidents and traffic volumes and produces reports for monitoring the evolution of key performance indicators for road safety, supporting decisions on actions for risk mitigation and safety improvements for road users. The paper illustrates the different types of analyses performed by the system. Pattern based analysis is used to evaluate safety performance indicators for the road sections matching defined patterns. Two different road segmentation algorithms, used to identify the most critical road sections according to various severity indicators, are presented and discussed. Differential analysis compares the value of selected severity indicators before and after the implementation of an intervention on a road. Finally, a graphical user interface allows the accident locations to be visualized and accidents with specific characteristics to be highlighted. The system was evaluated on the data collected between 2009 and 2011 for the A15 motorway in Italy, connecting Parma to La Spezia.

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.

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.

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.

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.

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...

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.

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.

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

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.

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.

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.

Cognitive systems engineering analysis of the JCO criticality accident

International Nuclear Information System (INIS)

Tanabe, Fumiya; Yamaguchi, Yukichi

2000-01-01

The JCO Criticality Accident is analyzed with a framework based on cognitive systems engineering. With the framework, analysis is conducted integrally both from the system viewpoint and actors viewpoint. The occupational chemical risk was important as safety constraint for the actors as well as the nuclear risk, which is due to criticality accident, to the public and to actors. The inappropriate actor's mental model of the work system played a critical role and several factors (e.g. poor training and education, lack of information on criticality safety control in the procedures and instructions, and lack of warning signs at workplace) contributed to form and shape the mental model. Based on the analysis, several countermeasures, such as warning signs, information system for supporting actors and improved training and education, are derived to prevent such an accident. (author)

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

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.))

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.

Systemic accident analysis: examining the gap between research and practice.

Science.gov (United States)

Underwood, Peter; Waterson, Patrick

2013-06-01

The systems approach is arguably the dominant concept within accident analysis research. Viewing accidents as a result of uncontrolled system interactions, it forms the theoretical basis of various systemic accident analysis (SAA) models and methods. Despite the proposed benefits of SAA, such as an improved description of accident causation, evidence within the scientific literature suggests that these techniques are not being used in practice and that a research-practice gap exists. The aim of this study was to explore the issues stemming from research and practice which could hinder the awareness, adoption and usage of SAA. To achieve this, semi-structured interviews were conducted with 42 safety experts from ten countries and a variety of industries, including rail, aviation and maritime. This study suggests that the research-practice gap should be closed and efforts to bridge the gap should focus on ensuring that systemic methods meet the needs of practitioners and improving the communication of SAA research. Copyright © 2013 Elsevier Ltd. All rights reserved.

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)

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

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.

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.

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

ADAM: An Accident Diagnostic,Analysis and Management System - Applications to Severe Accident Simulation and Management

International Nuclear Information System (INIS)

Zavisca, M.J.; Khatib-Rahbar, M.; Esmaili, H.; Schulz, R.

2002-01-01

The Accident Diagnostic, Analysis and Management (ADAM) computer code has been developed as a tool for on-line applications to accident diagnostics, simulation, management and training. ADAM's severe accident simulation capabilities incorporate a balance of mechanistic, phenomenologically based models with simple parametric approaches for elements including (but not limited to) thermal hydraulics; heat transfer; fuel heatup, meltdown, and relocation; fission product release and transport; combustible gas generation and combustion; and core-concrete interaction. The overall model is defined by a relatively coarse spatial nodalization of the reactor coolant and containment systems and is advanced explicitly in time. The result is to enable much faster than real time (i.e., 100 to 1000 times faster than real time on a personal computer) applications to on-line investigations and/or accident management training. Other features of the simulation module include provision for activation of water injection, including the Engineered Safety Features, as well as other mechanisms for the assessment of accident management and recovery strategies and the evaluation of PSA success criteria. The accident diagnostics module of ADAM uses on-line access to selected plant parameters (as measured by plant sensors) to compute the thermodynamic state of the plant, and to predict various margins to safety (e.g., times to pressure vessel saturation and steam generator dryout). Rule-based logic is employed to classify the measured data as belonging to one of a number of likely scenarios based on symptoms, and a number of 'alarms' are generated to signal the state of the reactor and containment. This paper will address the features and limitations of ADAM with particular focus on accident simulation and management. (authors)

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

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)

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)

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)

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

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

Analysis of occupational accidents with biological material among professionals in pre-hospital services

OpenAIRE

Oliveira,Adriana Cristina de; Paiva,Maria Henriqueta Rocha Siqueira

2013-01-01

OBJECTIVE: To estimate the prevalence of accidents due to biological material exposure, the characteristics and post-accident conduct among professionals of pre-hospital services of the four municipalities of Minas Gerais, Brazil. METHOD: A cross-sectional study, using a structured questionnaire that was developed to enable the calculation of prevalence, descriptive analysis and analytical analysis using logistic regression. The study included 228 professionals; the prevalence of accidents du...

Accident Damage Analysis Module (ADAM) – Technical Guidance, Software tool for Consequence Analysis calculations

OpenAIRE

FABBRI LUCIANO; BINDA MASSIMO; BRUINEN DE BRUIN YURI

2017-01-01

This report provides a technical description of the modelling and assumptions of the Accident Damage Analysis Module (ADAM) software application, which has been recently developed by the Joint Research Centre (JRC) of the European Commission (EC) to assess physical effects of an industrial accident resulting from an unintended release of a dangerous substance

Statistical analysis of accident data associated with sea transport (invited paper)

International Nuclear Information System (INIS)

Raffestin, D.; Armingaud, F.; Schneider, T.; Delaigue, S.

1998-01-01

This analysis, based on Lloyd's database, gives an accurate description of the world fleet and the most severe ship accidents, as well as the frequencies of accident per ship type, accident category and age category. Complementary analyses were achieved using fire accident databases from AEA Technology and the French Bureau Veritas. The results should be used in the perspective of safety assessments of maritime shipments of radioactive material. For this purpose the existence of the regulations of the International Maritime Organisation has to be considered, leading to the introduction of correction factors to these statistical data derived from general cargo-carrying ships. (author)

Estimating the causes of traffic accidents using logistic regression and discriminant analysis.

Science.gov (United States)

Karacasu, Murat; Ergül, Barış; Altin Yavuz, Arzu

2014-01-01

Factors that affect traffic accidents have been analysed in various ways. In this study, we use the methods of logistic regression and discriminant analysis to determine the damages due to injury and non-injury accidents in the Eskisehir Province. Data were obtained from the accident reports of the General Directorate of Security in Eskisehir; 2552 traffic accidents between January and December 2009 were investigated regarding whether they resulted in injury. According to the results, the effects of traffic accidents were reflected in the variables. These results provide a wealth of information that may aid future measures toward the prevention of undesired results.

Computational analysis of the behaviour of nuclear fuel under steady state, transient and accident conditions

International Nuclear Information System (INIS)

2007-12-01

Accident analysis is an important tool for ensuring the adequacy and efficiency of the provision in the defence in depth concept to cope with challenges to plant safety. Accident analysis is the milestone of the demonstration that the plant is capable of meeting any prescribed limits for radioactive releases and any other acceptable limits for the safe operation of the plant. It is used, by designers, utilities and regulators, in a number of applications such as: (a) licensing of new plants, (b) modification of existing plants, (c) analysis of operational events, (d) development, improvement or justification of the plant operational limits and conditions, and (e) safety cases. According to the defence in depth concept, the fuel rod cladding constitutes the first containment barrier of the fission products. Therefore, related safety objectives and associated criteria are defined, in order to ensure, at least for normal operation and anticipated transients, the integrity of the cladding, and for accident conditions, acceptable radiological consequences with regard to the postulated frequency of the accident, as usually identified in the safety analysis reports. Therefore, computational analysis of fuel behaviour under steady state, transient and accident conditions constitutes a major link of the safety case in order to justify the design and the safety of the fuel assemblies, as far as all relevant phenomena are correctly addressed and modelled. This publication complements the IAEA Safety Report on Accident Analysis for Nuclear Power Plants (Safety Report Series No. 23) that provides practical guidance for establishing a set of conceptual and formal methods and practices for performing accident analysis. Computational analysis of the behaviour of nuclear fuel under transient and accident conditions, including normal operation (e.g. power ramp rates) is developed in this publication. For design basis accidents, depending on the type of influence on a fuel element

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)

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

Statistical analysis of accident data associated with sea transport (data from 1994-1997). Annex 1

International Nuclear Information System (INIS)

Schneider, T.; Tabarre, M.; Armingaud, F.

2001-01-01

This analysis is based on Lloyd's database concerning sea transport accidents for the 1994-1997 period and completes the previous analysis based on 1994 data. It gives an accurate description of the world fleet and the most severe ship accidents (total losses), as well as the frequencies of accident (in average on the 1994-1997 period the frequency of accident for cargo carrying ships is 2.57.10 -3 loss /ship.year). Furthermore, an analysis has been performed on the ship casualties recorded by the Marine Accident Investigation Branch (MAIB) for UK vessels for the 1990-1996 period, this database including all accidents for which a declaration has been made to authorities (for example, the average frequency of fires derived from this analysis is 1.36.10 -2 per ship.year, this occurrence corresponding to the occurrence of initiating events of fire). Concerning fire accidents aboard ships supposed to be representative of the radioactive material transporters, a specific analysis was achieved by the French Bureau Veritas, on a selection of the world casualties (total losses) for the 1978-1988 period. This analysis related to the origin of the fire points out that it originates mainly in the machinery room and quarters. In a few cases the fire duration recorded is more than one day. (author)

[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.

Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

International Nuclear Information System (INIS)

O'Kula, K.R.; Chung, D.Y.

1999-01-01

Since its inception, the Accident Analysis Subgroup (AAS) of the Energy Facility Contractors Group (EFCOG) has been a leading organization promoting development and application of appropriate methodologies for safety analysis of US Department of Energy (DOE) installations. The AAS, one of seven chartered by the EFCOG Safety Analysis Working Group, has performed an oversight function and provided direction to several technical groups. These efforts have been instrumental toward formal evaluation of computer models, improving the pedigree on high-use computer models, and development of the user-friendly Accident Analysis Guidebook (AAG). All of these improvements have improved the analytical toolkit for best complying with DOE orders and standards shaping safety analysis reports (SARs) and related documentation. Major support for these objectives has been through DOE/DP-45

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

The Driver Behaviour Questionnaire as accident predictor; A methodological re-meta-analysis.

Science.gov (United States)

Af Wåhlberg, A E; Barraclough, P; Freeman, J

2015-12-01

The Manchester Driver Behaviour Questionnaire (DBQ) is the most commonly used self-report tool in traffic safety research and applied settings. It has been claimed that the violation factor of this instrument predicts accident involvement, which was supported by a previous meta-analysis. However, that analysis did not test for methodological effects, or include unpublished results. The present study re-analysed studies on prediction of accident involvement from DBQ factors, including lapses, and many unpublished effects. Tests of various types of dissemination bias and common method variance were undertaken. Outlier analysis showed that some effects were probably not reliable data, but excluding them did not change the results. For correlations between violations and crashes, tendencies for published effects to be larger than unpublished ones and for effects to decrease over time were observed, but were not significant. Also, using the mean of accidents as proxy for effect indicated that studies where effects for violations are not reported have smaller effect sizes. These differences indicate dissemination bias. Studies using self-reported accidents as dependent variables had much larger effects than those using recorded accident data. Also, zero-order correlations were larger than partial correlations controlled for exposure. Similarly, violations/accidents effects were strong only when there was also a strong correlation between accidents and exposure. Overall, the true effect is probably very close to zero (rresearch. Also, validation of self-reports should be more comprehensive in the future, taking into account the possibility of common method variance. Copyright © 2015 Elsevier Ltd and National Safety Council. All rights reserved.

Domino effect in chemical accidents: main features and accident sequences.

Science.gov (United States)

Darbra, R M; Palacios, Adriana; Casal, Joaquim

2010-11-15

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 are external events (31%) and mechanical failure (29%). Storage areas (35%) and process plants (28%) are by far the most common settings for domino accidents. Eighty-nine per cent of the accidents involved flammable materials, the most frequent of which was LPG. The domino effect sequences were analyzed using relative probability event trees. The most frequent sequences were explosion→fire (27.6%), fire→explosion (27.5%) and fire→fire (17.8%). Copyright © 2010 Elsevier B.V. All rights reserved.

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

Science.gov (United States)

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

2016-01-01

To guarantee the security of water quality in water transfer channels, especially in open channels, analysis of potential emergent pollution sources in the water transfer process is critical. It is also indispensable for forewarnings and protection from emergent pollution accidents. Bridges above open channels with large amounts of truck traffic are the main locations where emergent accidents could occur. A Bayesian Network model, which consists of six root nodes and three middle layer nodes, was developed in this paper, and was employed to identify the possibility of potential pollution risk. Dianbei Bridge is reviewed as a typical bridge on an open channel of the Middle Route of the South to North Water Transfer Project where emergent traffic accidents could occur. Risk of water pollutions caused by leakage of pollutants into water is focused in this study. The risk for potential traffic accidents at the Dianbei Bridge implies a risk for water pollution in the canal. Based on survey data, statistical analysis, and domain specialist knowledge, a Bayesian Network model was established. The human factor of emergent accidents has been considered in this model. Additionally, this model has been employed to describe the probability of accidents and the risk level. The sensitive reasons for pollution accidents have been deduced. The case has also been simulated that sensitive factors are in a state of most likely to lead to accidents. Copyright © 2015 Elsevier Ltd. All rights reserved.

Accident Analysis Guidance for Completion of 10 CFR 830-Compliant DSAs

International Nuclear Information System (INIS)

Vincent, A.

2002-01-01

Safety analysis contractors responsible for existing nuclear facilities are required to submit a Documented Safety Analysis to the Department of Energy for approval by April 2003. Recognizing that schedule and resource limitations may be significant, an initiative is underway to make available a set of guidance tools. The guidance is in the form of a peer-reviewed Accident Analysis Guidebook, a series of application guides for ''safe harbor'' computer codes, establishment of a configuration-controlled collection of safety analysis software and a central registry to maintain it, and periodic analytical training on accident analysis methods. Delivery of the majority of these products is scheduled to be in FY 2003

Challenging the immediate causes: A work accident investigation in an oil refinery using organizational analysis.

Science.gov (United States)

Beltran, Sandra Lorena; Vilela, Rodolfo Andrade de Gouveia; de Almeida, Ildeberto Muniz

2018-01-01

In many companies, investigations of accidents still blame the victims without exploring deeper causes. Those investigations are reactive and have no learning potential. This paper aims to debate the historical organizational aspects of a company whose policy was incubating an accident. The empirical data are analyzed as part of a qualitative study of an accident that occurred in an oil refinery in Brazil in 2014. To investigate and analyse this case we used one-to-one and group interviews, participant observation, Collective Analyses of Work and a documentary review. The analysis was conducted on the basis of concepts of the Organizational Analysis of the event and the Model for Analysis and Prevention of Work Accidents. The accident had its origin in the interaction of social and organizational factors, among them being: excessively standardized culture, management tools and outcome indicators that give a false sense of safety, the decision to speed up the project, the change of operator to facilitate this outcome and performance management that encourages getting around the usual barriers. The superficial accident analysis conducted by the company that ignored human and organizational factors reinforces the traditional safety culture and favors the occurrence of new accidents.

Theories of radiation effects and reactor accident analysis

International Nuclear Information System (INIS)

Williams, P.M.; Ball, S.J.

1996-01-01

Muckerheide's paper was a public breakthrough on how one might assess the public health effects of low-level radiation. By the organization of a wealth of data, including the consequences of Hiroshima and Nagasaki but not including Chernobyl, he was able to conclude that present radioactive waste disposal and cleanup efforts need to be much less arduous than forecast by the U.S. Department of Energy, which, together with regulators, uses the linear hypothesis of radiation damage to humans. While the linear hypothesis is strongly defended and even recommended for extension to noncarcinogenic pollutants, exploration of a conservative threshold for very low level exposures could save billions of dollars in disposing of radioactive waste, enhance the understanding of reactor accident consequences, and assist in the development of design and operating criteria pertaining to severe accidents. In this context, the authors discuss the major differences between design-basis and severe accidents. The authors propose that what should ultimately be done is to develop a regulatory formula for severe-accident analysis that relates the public health effects to the amount and type of radionuclides released and distributed by the Chernobyl accident. Answers to the following important questions should provide the basis of this study: (1) What should be the criteria for distinguishing between design-basis and severe accidents, and what should be the basis for these criteria? (2) How do, and should, these criteria differ for older plants, newer operating plants, type of plant (i.e., gas cooled, water cooled, and liquid metal), advanced designs, and plants of the former Soviet Union? (3) How safe is safe enough?

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)

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

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

Analysis of Hydrogen Control Strategy Using Igniter during Severe Accident

International Nuclear Information System (INIS)

Lee, Sung Bok; Kim, Hyeong Taek; Lee, Keo Hyoung

2008-01-01

The Severe Accident Management Guidelines (SAMGs) for the operating pressurized water reactor (PWR) have been completed within 2006. Among the SAMG strategies, mitigation-07 is the most important strategy for managing a severe accident of a PWR in order to reduce containment hydrogen. The fastest way to reduce the containment hydrogen concentration is to intentionally ignite the hydrogen. For this strategy, igniters exist in Optimized Power Reactor 1000 (OPR 1000) to burn hydrogen for a severe accident. For using the igniters during a severe accident, the adverse effects such as the explosion of the hydrogen mixture should be considered for containment integrity. However, an applicable discrimination method to activate the igniters does not exist, so that the hydrogen control strategy using the igniters cannot be chosen during a severe accident. Thus, this study focused on suggesting an applicable discrimination method to carry out the strategy of using the igniters. In this study, the specific plant used for this analysis is Ulchin Unit 5 and 6, OPR 1000 plant, in Korea

Introduction of Bayesian network in risk analysis of maritime accidents in Bangladesh

Science.gov (United States)

Rahman, Sohanur

2017-12-01

Due to the unique geographic location, complex navigation environment and intense vessel traffic, a considerable number of maritime accidents occurred in Bangladesh which caused serious loss of life, property and environmental contamination. Based on the historical data of maritime accidents from 1981 to 2015, which has been collected from Department of Shipping (DOS) and Bangladesh Inland Water Transport Authority (BIWTA), this paper conducted a risk analysis of maritime accidents by applying Bayesian network. In order to conduct this study, a Bayesian network model has been developed to find out the relation among parameters and the probability of them which affect accidents based on the accident investigation report of Bangladesh. Furthermore, number of accidents in different categories has also been investigated in this paper. Finally, some viable recommendations have been proposed in order to ensure greater safety of inland vessels in Bangladesh.

A human factors analysis of fatal and serious injury accidents in Alaska, 2004-2009.

Science.gov (United States)

2011-12-01

"This report summarizes the analysis of 97 general aviation accidents in Alaska that resulted in a fatality or serious : injury to one or more aircraft occupants for the years 2004-2009. The accidents were analyzed using the Human : Factors Analysis ...

Work-related accidents among the Iranian population: a time series analysis, 2000-2011.

Science.gov (United States)

Karimlou, Masoud; Salehi, Masoud; Imani, Mehdi; Hosseini, Agha-Fatemeh; Dehnad, Afsaneh; Vahabi, Nasim; Bakhtiyari, Mahmood

2015-01-01

Work-related accidents result in human suffering and economic losses and are considered as a major health problem worldwide, especially in the economically developing world. To introduce seasonal autoregressive moving average (ARIMA) models for time series analysis of work-related accident data for workers insured by the Iranian Social Security Organization (ISSO) between 2000 and 2011. In this retrospective study, all insured people experiencing at least one work-related accident during a 10-year period were included in the analyses. We used Box-Jenkins modeling to develop a time series model of the total number of accidents. There was an average of 1476 accidents per month (1476·05±458·77, mean±SD). The final ARIMA (p,d,q) (P,D,Q)s model for fitting to data was: ARIMA(1,1,1)×(0,1,1)12 consisting of the first ordering of the autoregressive, moving average and seasonal moving average parameters with 20·942 mean absolute percentage error (MAPE). The final model showed that time series analysis of ARIMA models was useful for forecasting the number of work-related accidents in Iran. In addition, the forecasted number of work-related accidents for 2011 explained the stability of occurrence of these accidents in recent years, indicating a need for preventive occupational health and safety policies such as safety inspection.

[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.

Preliminary Assessment of ICRP Dose Conversion Factor Recommendations for Accident Analysis Applications

International Nuclear Information System (INIS)

Vincent, A.M.

2002-01-01

Accident analysis for U.S. Department of Energy (DOE) nuclear facilities is an integral part of the overall safety basis developed by the contractor to demonstrate facility operation can be conducted safely. An appropriate documented safety analysis for a facility discusses accident phenomenology, quantifies source terms arising from postulated process upset conditions, and applies a standardized, internationally-recognized database of dose conversion factors (DCFs) to evaluate radiological conditions to offsite receptors

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

ACCIDENT ANALYSES & CONTROL OPTIONS IN SUPPORT OF THE SLUDGE WATER SYSTEM SAFETY ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

WILLIAMS, J.C.

2003-11-15

This report documents the accident analyses and nuclear safety control options for use in Revision 7 of HNF-SD-WM-SAR-062, ''K Basins Safety Analysis Report'' and Revision 4 of HNF-SD-SNF-TSR-001, ''Technical Safety Requirements - 100 KE and 100 KW Fuel Storage Basins''. These documents will define the authorization basis for Sludge Water System (SWS) operations. This report follows the guidance of DOE-STD-3009-94, ''Preparation Guide for US. Department of Energy Nonreactor Nuclear Facility Safety Analysis Reports'', for calculating onsite and offsite consequences. The accident analysis summary is shown in Table ES-1 below. While this document describes and discusses potential control options to either mitigate or prevent the accidents discussed herein, it should be made clear that the final control selection for any accident is determined and presented in HNF-SD-WM-SAR-062.

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

Application of Latin hypercube sampling to RADTRAN 4 truck accident risk sensitivity analysis

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.; Kanipe, F.L.

1994-01-01

The sensitivity of calculated dose estimates to various RADTRAN 4 inputs is an available output for incident-free analysis because the defining equations are linear and sensitivity to each variable can be calculated in closed mathematical form. However, the necessary linearity is not characteristic of the equations used in calculation of accident dose risk, making a similar tabulation of sensitivity for RADTRAN 4 accident analysis impossible. Therefore, a study of sensitivity of accident risk results to variation of input parameters was performed using representative routes, isotopic inventories, and packagings. It was determined that, of the approximately two dozen RADTRAN 4 input parameters pertinent to accident analysis, only a subset of five or six has significant influence on typical analyses or is subject to random uncertainties. These five or six variables were selected as candidates for Latin Hypercube Sampling applications. To make the effect of input uncertainties on calculated accident risk more explicit, distributions and limits were determined for two variables which had approximately proportional effects on calculated doses: Pasquill Category probability (PSPROB) and link population density (LPOPD). These distributions and limits were used as input parameters to Sandia's Latin Hypercube Sampling code to generate 50 sets of RADTRAN 4 input parameters used together with point estimates of other necessary inputs to calculate 50 observations of estimated accident dose risk.Tabulations of the RADTRAN 4 accident risk input variables and their influence on output plus illustrative examples of the LHS calculations, for truck transport situations that are typical of past experience, will be presented

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

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.

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.

Spatial Analysis of Accident Spots Using Weighted Severity Index ...

African Journals Online (AJOL)

ADOWIE PERE

Spatial Analysis of Accident Spots Using Weighted Severity Index (WSI) and ... pedestrians avoiding the use of pedestrian bridges/aid even when they are available. ..... not minding an unforeseen obstruction, miscalculations and wrong break.

Accident analysis of railway transportation of low-level radioactive and hazardous chemical wastes: Application of the /open quotes/Maximum Credible Accident/close quotes/ concept

Energy Technology Data Exchange (ETDEWEB)

Ricci, E.; McLean, R.B.

1988-09-01

The maximum credible accident (MCA) approach to accident analysis places an upper bound on the potential adverse effects of a proposed action by using conservative but simplifying assumptions. It is often used when data are lacking to support a more realistic scenario or when MCA calculations result in acceptable consequences. The MCA approach can also be combined with realistic scenarios to assess potential adverse effects. This report presents a guide for the preparation of transportation accident analyses based on the use of the MCA concept. Rail transportation of contaminated wastes is used as an example. The example is the analysis of the environmental impact of the potential derailment of a train transporting a large shipment of wastes. The shipment is assumed to be contaminated with polychlorinated biphenyls and low-level radioactivities of uranium and technetium. The train is assumed to plunge into a river used as a source of drinking water. The conclusions from the example accident analysis are based on the calculation of the number of foreseeable premature cancer deaths the might result as a consequence of this accident. These calculations are presented, and the reference material forming the basis for all assumptions and calculations is also provided.

Accident analysis of railway transportation of low-level radioactive and hazardous chemical wastes: Application of the /open quotes/Maximum Credible Accident/close quotes/ concept

International Nuclear Information System (INIS)

Ricci, E.; McLean, R.B.

1988-09-01

The maximum credible accident (MCA) approach to accident analysis places an upper bound on the potential adverse effects of a proposed action by using conservative but simplifying assumptions. It is often used when data are lacking to support a more realistic scenario or when MCA calculations result in acceptable consequences. The MCA approach can also be combined with realistic scenarios to assess potential adverse effects. This report presents a guide for the preparation of transportation accident analyses based on the use of the MCA concept. Rail transportation of contaminated wastes is used as an example. The example is the analysis of the environmental impact of the potential derailment of a train transporting a large shipment of wastes. The shipment is assumed to be contaminated with polychlorinated biphenyls and low-level radioactivities of uranium and technetium. The train is assumed to plunge into a river used as a source of drinking water. The conclusions from the example accident analysis are based on the calculation of the number of foreseeable premature cancer deaths the might result as a consequence of this accident. These calculations are presented, and the reference material forming the basis for all assumptions and calculations is also provided

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...

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

Analysis of Human Errors in Industrial Incidents and Accidents for Improvement of Work Safety

DEFF Research Database (Denmark)

Leplat, J.; Rasmussen, Jens

1984-01-01

Methods for the analysis of work accidents are discussed, and a description is given of the use of a causal situation analysis in terms of a 'variation tree' in order to explain the course of events of the individual cases and to identify possible improvements. The difficulties in identifying...... 'causes' of accidents are discussed, and it is proposed to analyze accident reports with the specific aim of identifying the potential for future improvements rather than causes of past events. In contrast to traditional statistical analysis of work accident data, which typically give very general...... recommendations, the method proposed identifies very explicit countermeasures. Improvements require a change in human decisions during equipment design, work planning, or the execution itself. The use of a model of human behavior drawing a distinction between automated skill-based behavior, rule-based 'know...

Feasibility study on the rod ejection accident analysis with RETRAN-MASTER code system

International Nuclear Information System (INIS)

Kim, Y. H.; Lee, C. S.

2003-01-01

KEPRI has been developed the in-house methodology for non-LOCA safety analyses based on the codes and methodologies of vendors and EPRI. Using the methodology, the rod ejection accident, which is classified into the generic accident analysis category of reactivity insertion accident in primary system, has been analyzed with RETRAN-MASTER code system. And the feasibility of the coupled code system has been verified by the review of the results. Furthermore, to assess the important parameters to the accident, the sensitivity analyses have been carried out over some parameters

A flammability and combustion model for integrated accident analysis

International Nuclear Information System (INIS)

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

1988-01-01

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

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)

How perceptions of experience-based analysis influence explanations of work accidents.

Science.gov (United States)

Mbaye, Safiétou; Kouabenan, Dongo Rémi

2013-12-01

This article looks into how perceptions of experience-based analysis (EBA) influence causal explanations of accidents given by managers and workers in the chemical industry (n=409) and in the nuclear industry (n=222). The approach is based on the model of naive explanations of accidents (Kouabenan, 1999, 2006, 2009), which recommends taking into account explanations of accidents spontaneously given by individuals, including laypersons, not only to better understand why accidents occur but also to design and implement the most appropriate prevention measures. The study reported here describes the impact of perceptions about EBA (perceived effectiveness, personal commitment, and the feeling of being involved in EBA practices) on managers' and workers' explanations of accidents likely to occur at the workplace. The results indicated that both managers and workers made more internal explanations than external ones when they perceived EBA positively. Moreover, the more the participants felt involved in EBA, were committed to it, and judged it effective, the more they explained accidents in terms of factors internal to the workers. Recommendations are proposed for reducing defensive reactions, increasing personal commitment to EBA, and improving EBA effectiveness. © 2013.

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

Fuel relocation modeling in the SAS4A accident analysis code system

International Nuclear Information System (INIS)

Tentner, A.M.; Miles, K.J.; Kalimullah; Hill, D.J.

1986-01-01

The SAS4A code system has been designed for the analysis of the initial phase of Hypothetical Core Disruptive Accidents (HCDAs) up to gross melting or failure of the subassembly walls. During such postulated accident scenarios as the Loss-of-Flow (LOF) and Transient-Overpower (TOP) events, the relocation of the fuel plays a key role in determining the sequence of events and the amount of energy produced before neutronic shutdown. This paper discusses the general strategy used in modelong the various phenomena which lead to fuel relocation and presents the key fuel relocation models used in SAS4A. The implications of these models for the whole-core accident analysis as well as recent results of fuel relocation are emphasized. 12 refs

Media content analysis of the Fukushima accident in two Belgian newspapers

International Nuclear Information System (INIS)

Perko, T.; Turcanu, C.; Geenen, D.; Mamane, N.; Van Rooy, L.

2011-01-01

In case of a nuclear accident, the media play a major role in communicating with the public. It is therefore crucial to know what messages are the media delivering in a nuclear emergency and how do they frame the event. Analysing the media reporting on the Fukushima nuclear accident can benefit nuclear emergency management in two major aspects. On the one hand, such analysis shows how to deliver risk messages effectively through the media and on the other hand, it brings insights into the information that has to be communicated by the emergency managers to the mass media. The media analysis of the nuclear accident in Fukushima reported here was done by means of discourse and content analysis. The coding method followed explicit rules of coding and enabled large quantities of data to be categorized. The newspapers included in the analysis were the Belgian newspapers Le Soir (French language) and De Standaard (Dutch language). The media news were obtained from press clippings by Media data base at University Antwerp - MEDIARGUS for the period between 11th of March to 11th of May, 2011.

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

Energy Technology Data Exchange (ETDEWEB)

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

2011-06-01

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

Solid waste accident analysis in support of the Savannah River Waste Management Environmental Impact Statement

International Nuclear Information System (INIS)

Copeland, W.J.; Crumm, A.T.; Kearnaghan, D.P.; Rabin, M.S.; Rossi, D.E.

1994-07-01

The potential for facility accidents and the magnitude of their impacts are important factors in the evaluation of the solid waste management addressed in the Environmental Impact Statement. The purpose of this document is to address the potential solid waste management facility accidents for comparative use in support of the Environmental Impact Statement. This document must not be construed as an Authorization Basis document for any of the SRS waste management facilities. Because of the time constraints placed on preparing this accident impact analysis, all accident information was derived from existing safety documentation that has been prepared for SRS waste management facilities. A list of facilities to include in the accident impact analysis was provided as input by the Savannah River Technology Section. The accident impact analyses include existing SRS waste management facilities as well as proposed facilities. Safety documentation exists for all existing and many of the proposed facilities. Information was extracted from this existing documentation for this impact analysis. There are a few proposed facilities for which safety analyses have not been prepared. However, these facilities have similar processes to existing facilities and will treat, store, or dispose of the same type of material that is in existing facilities; therefore, the accidents can be expected to be similar

Accidents at work and costs analysis: a field study in a large Italian company.

Science.gov (United States)

Battaglia, Massimo; Frey, Marco; Passetti, Emilio

2014-01-01

Accidents at work are still a heavy burden in social and economic terms, and action to improve health and safety standards at work offers great potential gains not only to employers, but also to individuals and society as a whole. However, companies often are not interested to measure the costs of accidents even if cost information may facilitate preventive occupational health and safety management initiatives. The field study, carried out in a large Italian company, illustrates technical and organisational aspects associated with the implementation of an accident costs analysis tool. The results indicate that the implementation (and the use) of the tool requires a considerable commitment by the company, that accident costs analysis should serve to reinforce the importance of health and safety prevention and that the economic dimension of accidents is substantial. The study also suggests practical ways to facilitate the implementation and the moral acceptance of the accounting technology.

Containment integrity analysis under accidents

International Nuclear Information System (INIS)

Lin Chengge; Zhao Ruichang; Liu Zhitao

2010-01-01

Containment integrity analyses for current nuclear power plants (NPPs) mainly focus on the internal pressure caused by design basis accidents (DBAs). In addition to the analyses of containment pressure response caused by DBAs, the behavior of containment during severe accidents (SAs) are also evaluated for AP1000 NPP. Since the conservatism remains in the assumptions,boundary conditions and codes, margin of the results of containment integrity analyses may be overestimated. Along with the improvements of the knowledge to the phenomena and process of relevant accidents, the margin overrated can be appropriately reduced by using the best estimate codes combined with the uncertainty methods, which could be beneficial to the containment design and construction of large passive plants (LPP) in China. (authors)

Analysis on the nitrogen drilling accident of Well Qionglai 1 (II: Restoration of the accident process and lessons learned

Directory of Open Access Journals (Sweden)

Yingfeng Meng

2015-12-01

Full Text Available All the important events of the accident of nitrogen drilling of Well Qionglai 1 have been speculated and analyzed in the paper I. In this paper II, based on the investigating information, the well log data and some calculating and simulating results, according to the analysis method of the fault tree of safe engineering, the every possible compositions, their possibilities and time schedule of the events of the accident of Well Qionglai 1 have been analyzed, the implications of the logging data have been revealed, the process of the accident of Well Qionglai 1 has been restored. Some important understandings have been obtained: the objective causes of the accident is the rock burst and the induced events form rock burst, the subjective cause of the accident is that the blooie pipe could not bear the flow burden of the clasts from rock burst and was blocked by the clasts. The blocking of blooie pipe caused high pressure in wellhead, the high pressure made the blooie pipe burst, natural gas came out and flared fire. This paper also thinks that the rock burst in gas drilling in fractured tight sandstone gas zone is objective and not avoidable, but the accidents induced from rock burst can be avoidable by improving the performance of the blooie pipe, wellhead assemblies and drilling tool accessories aiming at the downhole rock burst.

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

Safety analysis of RA reactor operation, I-III, Part III - Environmental effect of the maximum credible accident

International Nuclear Information System (INIS)

Raisic, N.

1963-02-01

Maximum credible accident at the RA reactor would consider release of fission products into the environment. This would result from fuel elements failure or meltdown due to loss of coolant. The analysis presented in this report assumes that the reactor was operating at nominal power at the moment of maximum possible accident. The report includes calculations of fission products activity at the moment of accident, total activity release during the accident, concentration of radioactive material in the air in the reactor neighbourhood, and the analysis of accident environmental effects

Work-related accidents among the Iranian population: a time series analysis, 2000–2011

Science.gov (United States)

Karimlou, Masoud; Imani, Mehdi; Hosseini, Agha-Fatemeh; Dehnad, Afsaneh; Vahabi, Nasim; Bakhtiyari, Mahmood

2015-01-01

Background Work-related accidents result in human suffering and economic losses and are considered as a major health problem worldwide, especially in the economically developing world. Objectives To introduce seasonal autoregressive moving average (ARIMA) models for time series analysis of work-related accident data for workers insured by the Iranian Social Security Organization (ISSO) between 2000 and 2011. Methods In this retrospective study, all insured people experiencing at least one work-related accident during a 10-year period were included in the analyses. We used Box–Jenkins modeling to develop a time series model of the total number of accidents. Results There was an average of 1476 accidents per month (1476·05±458·77, mean±SD). The final ARIMA (p,d,q) (P,D,Q)s model for fitting to data was: ARIMA(1,1,1)×(0,1,1)12 consisting of the first ordering of the autoregressive, moving average and seasonal moving average parameters with 20·942 mean absolute percentage error (MAPE). Conclusions The final model showed that time series analysis of ARIMA models was useful for forecasting the number of work-related accidents in Iran. In addition, the forecasted number of work-related accidents for 2011 explained the stability of occurrence of these accidents in recent years, indicating a need for preventive occupational health and safety policies such as safety inspection. PMID:26119774

Enhancing AP1000 reactor accident management capabilities for long term accidents

International Nuclear Information System (INIS)

Jiang Pingting; Liu Mengying; Duan Chengjie; Liao Yehong

2015-01-01

Passive safety actions are considered as main measures under severe accident in AP1000 power plant. However, risk is still existed. According to PSA, several probable scenarios for AP1000 nuclear power plant are analyzed in this paper with MAAP the severe accident analysis code. According to the analysis results, several deficiencies of AP1000 severe accident management are found. The long term cooling and containment depressurization capability for AP1000 power plant appear to be most important factors under such accidents. Then, several temporary strategies for AP1000 power plant are suggested, including PCCWST temporary water supply strategy after 72h, temporary injection strategy for IRWST, hydrogen relief action in fuel building, which would improve the safety of AP1000 power plant. At last, assessments of effectiveness for these strategies are performed, and the results are compared with analysis without these strategies. The comparisons showed that correct actions of these strategies would effectively prevent the accident process of AP1000 power plant. (author)

[Proposal of a method for collective analysis of work-related accidents in the hospital setting].

Science.gov (United States)

Osório, Claudia; Machado, Jorge Mesquita Huet; Minayo-Gomez, Carlos

2005-01-01

The article presents a method for the analysis of work-related accidents in hospitals, with the double aim of analyzing accidents in light of actual work activity and enhancing the vitality of the various professions that comprise hospital work. This process involves both research and intervention, combining knowledge output with training of health professionals, fostering expanded participation by workers in managing their daily work. The method consists of stimulating workers to recreate the situation in which a given accident occurred, shifting themselves to the position of observers of their own work. In the first stage of analysis, workers are asked to show the work analyst how the accident occurred; in the second stage, the work accident victim and analyst jointly record the described series of events in a diagram; in the third, the resulting record is re-discussed and further elaborated; in the fourth, the work accident victim and analyst evaluate and implement measures aimed to prevent the accident from recurring. The article concludes by discussing the method's possibilities and limitations in the hospital setting.

Analysis of traffic accidents in children

Directory of Open Access Journals (Sweden)

Pavlekić Snežana

2006-01-01

Full Text Available Introduction: Violent health damages of different origin (accidents, murders, suicides in children and youth are one of the main causes of death and disabilities in this group of population in most countries. Objective: Objective of our paper was to analyze all related factors of traffic accidents involving children and to propose adequate measures of their prevention. Method: The analysis of fatal traffic accidents of children and youth aged to 18 years on the territory of Belgrade, within the period from 1998 to 2002. Results: In relation to other forms of violent death, the traffic mortality rate in children and youth holds the leading position, accounting for 56.9% with pedestrians as the most frequent category (57.4%. The most frequent age was between 7 and 9 years (46.8% and the boys were more frequently injured than the girls. It was established that the majority of children (51.9% was either running across the street outside the pedestrian/ zebra crossings or they were carelessly running out in the street, especially in April, July, August and September. More than a half of them (55.5%, predominantly school children, were injured by the end of working week, on Thursday and Friday. Conclusion: Results of our research have shown that the traffic education of children in our region is inadequate. Due to the abovementioned, it is primarily necessary to establish long-term and permanent education of this category of population. In addition, some public investments in the City infrastructure will be required in order to reduce the risk of traffic injuries in children.

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

Empirical Risk Analysis of Severe Reactor Accidents in Nuclear Power Plants after Fukushima

OpenAIRE

Kaiser, Jan Christian

2012-01-01

Many countries are reexamining the risks connected with nuclear power generation after the Fukushima accidents. To provide updated information for the corresponding discussion a simple empirical approach is applied for risk quantification of severe reactor accidents with International Nuclear and Radiological Event Scale (INES) level ≥5. The analysis is based on worldwide data of commercial nuclear facilities. An empirical hazard of 21 (95% confidence intervals (CI) 4; 62) severe accidents am...

Prevention of pedestrian accidents.

OpenAIRE

Kendrick, D

1993-01-01

Child pedestrian accidents are the most common road traffic accident resulting in injury. Much of the existing work on road traffic accidents is based on analysing clusters of accidents despite evidence that child pedestrian accidents tend to be more dispersed than this. This paper analyses pedestrian accidents in 573 children aged 0-11 years by a locally derived deprivation score for the years 1988-90. The analysis shows a significantly higher accident rate in deprived areas and a dose respo...

Radioactive material (RAM) accident/incident data analysis program

International Nuclear Information System (INIS)

Emerson, E.L.; McClure, J.D.

1985-03-01

This report describes the development of the Radioactive Material Transportation Accident/Incident Data Base (RAM-AIDB), which contains information on the occurrences of transportation accidents and incidents, for radioactive materials (RAM) that are involved in the process of transportation, loading and unloading operation, or temporary storage. These transportation operations are in support of the nuclear fuel cycle for electrical energy generation. This study analyzes in some detail basic accident/incident statistical data, RAM packaging accident response data, and the health effects associated with RAM transport accidents/incidents. This report presents a summary of US RAM transport accident/incident experience for the period 1971 through December 1981. In addition, a sample annual summary of accident/incident experience is presented for the calendar year 1981

A human factor analysis of a radiotherapy accident

International Nuclear Information System (INIS)

Thellier, S.

2009-01-01

Since September 2005, I.R.S.N. studies activities of radiotherapy treatment from the angle of the human and organizational factors to improve the reliability of treatment in radiotherapy. Experienced in nuclear industry incidents analysis, I.R.S.N. analysed and diffused in March 2008, for the first time in France, the detailed study of a radiotherapy accident from the angle of the human and organizational factors. The method used for analysis is based on interviews and documents kept by the hospital. This analysis aimed at identifying the causes of the difference recorded between the dose prescribed by the radiotherapist and the dose effectively received by the patient. Neither verbal nor written communication (intra-service meetings and protocols of treatment) allowed information to be transmitted correctly in order to permit radiographers to adjust the irradiation zones correctly. This analysis highlighted the fact that during the preparation and the carrying out of the treatment, various factors led planned controls to not be performed. Finally, this analysis highlighted the fact that unsolved areas persist in the report over this accident. This is due to a lack of traceability of a certain number of key actions. The article concluded that there must be improvement in three areas: cooperation between the practitioners, control of the actions and traceability of the actions. (author)

Accidents at Work and Costs Analysis: A Field Study in a Large Italian Company

Science.gov (United States)

BATTAGLIA, Massimo; FREY, Marco; PASSETTI, Emilio

2014-01-01

Accidents at work are still a heavy burden in social and economic terms, and action to improve health and safety standards at work offers great potential gains not only to employers, but also to individuals and society as a whole. However, companies often are not interested to measure the costs of accidents even if cost information may facilitate preventive occupational health and safety management initiatives. The field study, carried out in a large Italian company, illustrates technical and organisational aspects associated with the implementation of an accident costs analysis tool. The results indicate that the implementation (and the use) of the tool requires a considerable commitment by the company, that accident costs analysis should serve to reinforce the importance of health and safety prevention and that the economic dimension of accidents is substantial. The study also suggests practical ways to facilitate the implementation and the moral acceptance of the accounting technology. PMID:24869894

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

Accident analysis of heavy water cooled thorium breeder reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su’ud, Zaki; Takaki, Naoyuki

2015-01-01

power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition

Accident analysis of heavy water cooled thorium breeder reactor

Energy Technology Data Exchange (ETDEWEB)

Yulianti, Yanti [Department of Physics, University of Lampung Jl. Sumantri Brojonegoro No.1 Bandar Lampung, Indonesia Email: y-yanti@unila.ac.id (Indonesia); Su’ud, Zaki [Department of Physics, Bandung Institute of Technology Jl. Ganesha 10 Bandung, Indonesia Email: szaki@fi.itb.ac.id (Indonesia); Takaki, Naoyuki [Department of Nuclear Safety Engineering Cooperative Major in Nuclear Energy (Graduate School) 1-28-1 Tamazutsumi,Setagayaku, Tokyo158-8557, Japan Email: ntakaki@tcu.ac.jp (Japan)

2015-04-16

power reactor has a peak value before reactor has new balance condition. The analysis showed that temperatures of fuel and claddings during accident are still below limitations which are in secure condition.

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.

Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

International Nuclear Information System (INIS)

Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min

2016-01-01

As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI

Coupling Computer Codes for The Analysis of Severe Accident Using A Pseudo Shared Memory Based on MPI

Energy Technology Data Exchange (ETDEWEB)

Cho, Young Chul; Park, Chang-Hwan; Kim, Dong-Min [FNC Technology Co., Yongin (Korea, Republic of)

2016-10-15

As there are four codes in-vessel analysis code (CSPACE), ex-vessel analysis code (SACAP), corium behavior analysis code (COMPASS), and fission product behavior analysis code, for the analysis of severe accident, it is complex to implement the coupling of codes with the similar methodologies for RELAP and CONTEMPT or SPACE and CAP. Because of that, an efficient coupling so called Pseudo shared memory architecture was introduced. In this paper, coupling methodologies will be compared and the methodology used for the analysis of severe accident will be discussed in detail. The barrier between in-vessel and ex-vessel has been removed for the analysis of severe accidents with the implementation of coupling computer codes with pseudo shared memory architecture based on MPI. The remaining are proper choice and checking of variables and values for the selected severe accident scenarios, e.g., TMI accident. Even though it is possible to couple more than two computer codes with pseudo shared memory architecture, the methodology should be revised to couple parallel codes especially when they are programmed using MPI.

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

Improvement of severe accident analysis method for KSNP

Energy Technology Data Exchange (ETDEWEB)

Park, Jae Hong [Korea Institute of Nuclear Safety, Taejon (Korea, Republic of); Cho, Song Won; Cho, Youn Soo [Korea Radiation Technology Institute Co., Taejon (Korea, Republic of)

2002-03-15

The objective of this study is preparation of MELCOR 1.8.5 input deck for KSNP and simulation of some major severe accidents. The contents of this project are preparation of MELCOR 1.8.5 base input deck for KSNP to understand severe accident phenomena and to assess severe accident strategy, preparation of 20 cell containment input deck to simulate the distribution of hydrogen and fission products in containment, simulation of some major severe accident scenarios such as TLOFW, SBO, SBLOCA, MBLOCA, and LBLOCA. The method for MELCOR 1.8.5 input deck preparation can be used to prepare the input deck for domestic PWRs and to simulate severe accident experiments such as ISP-46. Information gained from analyses of severe accidents may be helpful to set up the severe accident management strategy and to develop regulatory guidance.

Analysis of occupational accidents with biological material among professionals in pre-hospital services.

Science.gov (United States)

de Oliveira, Adriana Cristina; Paiva, Maria Henriqueta Rocha Siqueira

2013-02-01

To estimate the prevalence of accidents due to biological material exposure, the characteristics and post-accident conduct among professionals of pre-hospital services of the four municipalities of Minas Gerais, Brazil. A cross-sectional study, using a structured questionnaire that was developed to enable the calculation of prevalence, descriptive analysis and analytical analysis using logistic regression. The study included 228 professionals; the prevalence of accidents due to biological material exposure was 29.4%, with 49.2% percutaneous, 10.4% mucousal, 6.0% non-intact skin, and 34.4% intact skin. Among the professionals injured, those that stood out were nursing technicians (41.9%) and drivers (28.3%). Notification of the occurrence of the accident occurred in 29.8% of the cases. Percutaneous exposure was associated with time of work in the organization (OR=2.51, 95% CI: 1.18 to 5.35, paccidents with biological material should be encouraged, along with professional evaluation/monitoring.

Analysis of nuclear accidents and associated problems relevant to public perception of risk

International Nuclear Information System (INIS)

Naschi, G.; Petrangeli, G.

1993-01-01

The analytical study of nuclear accidents, even if they are limited in number, forms a significant part of the vast discipline of industrial plant risk analysis. The retrospective analysis of the causes and various elements which contributed to the evolution of real accidents, as well as, the evaluation of the consequences and lessons learned, constitute a bank of information which, when suitably elaborated through a process of rational synthesis, can strongly influence the preparation of safety normatives, plant design specifications, environmental impacts assessments, and the perception of risk. This latter aspect is gaining importance today as growing public awareness and sensitivity towards the development and use of new technologies now bear heavily on new plant decision making. This paper examines how the public perception of risk regarding nuclear energy has been influenced by the events surrounding the Chernobyl and Three Mile Island accidents and the way in which information dissemination concerning these accidents was handled by mass media

Phenomenological uncertainty analysis of containment building pressure load caused by severe accident sequences

International Nuclear Information System (INIS)

Park, S.Y.; Ahn, K.I.

2014-01-01

Highlights: • Phenomenological uncertainty analysis has been applied to level 2 PSA. • The methodology provides an alternative to simple deterministic analyses and sensitivity studies. • A realistic evaluation provides a more complete characterization of risks. • Uncertain parameters of MAAP code for the early containment failure were identified. - Abstract: This paper illustrates an application of a severe accident analysis code, MAAP, to the uncertainty evaluation of early containment failure scenarios employed in the containment event tree (CET) model of a reference plant. An uncertainty analysis of containment pressure behavior during severe accidents has been performed for an optimum assessment of an early containment failure model. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences of a nuclear power plant. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to the in-vessel hydrogen generation, direct containment heating, and gas combustion. The basic approach of this methodology is to (1) develop severe accident scenarios for which containment pressure loads should be performed based on a level 2 PSA, (2) identify severe accident phenomena relevant to an early containment failure, (3) identify the MAAP input parameters, sensitivity coefficients, and modeling options that describe or influence the early containment failure phenomena, (4) prescribe the likelihood descriptions of the potential range of these parameters, and (5) evaluate the code predictions using a number of random combinations of parameter inputs sampled from the likelihood distributions

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

The issue of hypothetical whole-core accidents continues to play a significant role in assessment of the potential risk to the public associated with LMFBR operation in the USA. The paper briefly characterizes the changing nature of this role, with emphasis on the current risk-oriented perspective. It then describes the models and codes used for accident analysis in the USA which have been developed under DOE sponsorship and summarizes some specific applications of the codes to the current generation of fast reactors. An assessment of future trends in this area concludes the paper

Analysis of accident progression in the TEPCO Fukushima Daiichi Nuclear Power Station

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

One of the objectives of this study is to investigate the early stage of the TEPCO Fukushima Daiichi accident and to check the validity of the countermeasures against the accident. Last year the early stage of the accident was analyzed with use of RELAP5 code, and the longer term analysis was done by MELCOR code. This year, the simulation of reactor water level instrumentation behavior by MELCOR code was performed. Another objective of this study is to analyze of the long term cooling after the Fukushima Daiichi accident by TRACE5 code. In order to simulate the cooling conditions in Fukushima plants after the accident, the parametric calculations were done on the assumption of the existence of steam/liquid leak in Reactor Pressure Vessel (RPV) and Pressure Containment Vessel (PCV) and the variety of debris distribution in RPV and PCV. As a result, the debris distribution in RPV and PCV was estimated by referring plant parameter such as reactor pressure and temperature. (author)

Analysis and evaluation of the nuclear criticality accident in JCO CO. LTD in Japan

International Nuclear Information System (INIS)

Liu Hua; Liu Xinhua; Li Bing

2001-01-01

The author describes JCO criticality accident situation including the background, process chronology and emergency countermeasures taken of the accident and its radiation consequence. The analysis about the direct and root causes of the accident and some conclusions are also showed. The direct cause of the accident is the use of geometrically unsafe process equipment and personnel violation. However, the root cause is lack of efficient technical management. Therefore, it is necessary to emphasize the criticality safety in nuclear fuel cycle installations and enhance safety culture of regulatory and operational personnel

Classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel

International Nuclear Information System (INIS)

Wu Tao

1993-01-01

Based on the analysis of the difference between the accident severity categorization used in the Ministry of Railway and that used in the safety analysis of the transporting spent fuel, a method used for the classification of the railway accident in accordance with the requirement of the safety analysis of transporting spent fuel is suggested. The method classifies the railway accidents into 10 scenarios and make it possible to scale the accident through directly using the data documented by the Ministry of Railway without any additional effort

Analysis of media coverage and KINS communication activities on Fukushima accident

International Nuclear Information System (INIS)

Lee, Ki Hyung; Hwang, Sun Chul; Yun, Yuen Wha; Lee, Gye Hwi; Jeong, Jin A; Song, Hye Rim; Yang, Cho Hee

2012-01-01

The people and mass media of Korea, the closest country to Japan, showed great interest in Fukushima nuclear power plant accident. The Korean government and KINS (Korea Institute of Nuclear Safety) attempted to provide accurate information to the press through various communication actions. In this study, we conducted an in-depth analysis of the tendencies of the press according to the accident sequence and tracked the diffusion of this issue. The purpose of this study is to determine the properties of the crisis and essence of the issue. We also carry out a general evaluation and draw implications through an analysis of the communication actions of KINS

Enhanced Accident Tolerant Fuels for LWRS - A Preliminary Systems Analysis

Energy Technology Data Exchange (ETDEWEB)

Gilles Youinou; R. Sonat Sen

2013-09-01

The severe accident at Fukushima Daiichi nuclear plants illustrates the need for continuous improvements through developing and implementing technologies that contribute to safe, reliable and cost-effective operation of the nuclear fleet. Development of enhanced accident tolerant fuel contributes to this effort. These fuels, in comparison with the standard zircaloy – UO2 system currently used by the LWR industry, should be designed such that they tolerate loss of active cooling in the core for a longer time period (depending on the LWR system and accident scenario) while maintaining or improving the fuel performance during normal operations, operational transients, and design-basis events. This report presents a preliminary systems analysis related to most of these concepts. The potential impacts of these innovative LWR fuels on the front-end of the fuel cycle, on the reactor operation and on the back-end of the fuel cycle are succinctly described without having the pretension of being exhaustive. Since the design of these various concepts is still a work in progress, this analysis can only be preliminary and could be updated as the designs converge on their respective final version.

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

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.

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

Use of fuel failure correlations in accident analysis

International Nuclear Information System (INIS)

O'Dell, L.D.; Baars, R.E.; Waltar, A.E.

1975-05-01

The MELT-III code for analysis of a Transient Overpower (TOP) accident in an LMFBR is briefly described, including failure criteria currently applied in the code. Preliminary results of calculations exploring failure patterns in time and space in the reactor core are reported and compared for the two empirical fuel failure correlations employed in the code. (U.S.)

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

Analisis Kecelakaan Kerja Pada Proyek Bangunan Gedung (Analysis of Work Accident on Building Construction Projects)

OpenAIRE

Ferdiansyah, Deni; Winarno, Setya; M, Faisol A

2009-01-01

Work accidents and fatalities often happen in construction industry, thus a study on this matter to promote safety management needs a thorough analysis of their elements. The purpose of this paper is to identify various types of accidents, cost of accident and insurance premium, and also the correlation between the types of accidents and their costs. The data was collected from thirty construction projects around Daerah Istimewa Yogyakarta province and surrounding areas. These data included t...

Thermal-hydraulic analysis of the Three Mile Island Unit 2 reactor accident with THALES code

International Nuclear Information System (INIS)

Hashimoto, Kazuichiro; Soda, Kunihisa

1991-10-01

The OECD Nuclear Energy Agency (NEA) has established a Task Group in the Committee on the Safety of Nuclear Installations (CSNI) to perform an analysis of Three Mile Island Unit 2 (TMI-2) accident as a standard problem to benchmark severe accident computer codes and to assess the capability of the codes. The TMI-2 Analysis Exercise was performed at the Japan Atomic Energy Research Institute (JAERI) using the THALES (Thermal-Hydraulic Analysis of Loss-of-Coolant, Emergency Core Cooling and Severe Core Damage) - PM1/TMI code. The purpose of the analysis is to verify the capability of THALES-PM1/TMI code to describe accident progression in the actual plant. The present paper describes the final result of the TMI-2 Analysis Exercise performed at JAERI. (author)

Preliminary analysis of the transient overpower accident for CRBRP. Final report

International Nuclear Information System (INIS)

Kastenberg, W.E.; Frank, M.V.

1975-07-01

A preliminary analysis of the transient overpower accident for the Clinch River Breeder Reactor Plant (CRBRP) is presented. Several uncertainties in the analysis and the estimation of ramp rates during the transition to disassembly are discussed. The major conclusions are summarized

Application of accident progression event tree technology to the Savannah River Site Defense Waste Processing Facility SAR analysis

International Nuclear Information System (INIS)

Brandyberry, M.D.; Baker, W.H.; Wittman, R.S.; Amos, C.N.

1993-01-01

The Accident Analysis in the Safety Analysis Report (SAR) for the Savannah River Site (SRS) Defense Waste Processing Facility (DWPF) has recently undergone an upgrade. Non-reactor SARs at SRS (and other Department of Energy (DOE) sites) use probabilistic techniques to assess the frequency of accidents at their facilities. This paper describes the application of an extension of the Accident Progression Event Tree (APET) approach to accidents at the SRS DWPF. The APET technique allows an integrated model of the facility risk to be developed, where previous probabilistic accident analyses have been limited to the quantification of the frequency and consequences of individual accident scenarios treated independently. Use of an APET allows a more structured approach, incorporating both the treatment of initiators that are common to more than one accident, and of accident progression at the facility

NIF: Impacts of chemical accidents and comparison of chemical/radiological accident approaches

International Nuclear Information System (INIS)

Lazaro, M.A.; Policastro, A.J.; Rhodes, M.

1996-01-01

The US Department of Energy (DOE) proposes to construct and operate the National Ignition Facility (NIF). The goals of the NIF are to (1) achieve fusion ignition in the laboratory for the first time by using inertial confinement fusion (ICF) technology based on an advanced-design neodymium glass solid-state laser, and (2) conduct high-energy-density experiments in support of national security and civilian applications. The primary focus of this paper is worker-public health and safety issues associated with postulated chemical accidents during the operation of NIF. The key findings from the accident analysis will be presented. Although NIF chemical accidents will be emphasized, the important differences between chemical and radiological accident analysis approaches and the metrics for reporting results will be highlighted. These differences are common EIS facility and transportation accident assessments

Learning lessons from Natech accidents - the eNATECH accident database

Science.gov (United States)

Krausmann, Elisabeth; Girgin, Serkan

2016-04-01

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

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.

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)

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

Risk of Occupational Accidents in Workers with Obstructive Sleep Apnea: Systematic Review and Meta-analysis

Science.gov (United States)

Garbarino, Sergio; Guglielmi, Ottavia; Sanna, Antonio; Mancardi, Gian Luigi; Magnavita, Nicola

2016-01-01

Study Objectives: Obstructive sleep apnea (OSA) is the single most important preventable medical cause of excessive daytime sleepiness (EDS) and driving accidents. OSA may also adversely affect work performance through a decrease in productivity, and an increase in the injury rate. Nevertheless, no systematic review and meta-analysis of the relationship between OSA and work accidents has been performed thus far. Methods: PubMed, PsycInfo, Scopus, Web of Science, and Cochrane Library were searched. Out of an initial list of 1,099 papers, 10 studies (12,553 participants) were eligible for our review, and 7 of them were included in the meta-analysis. The overall effects were measured by odds ratios (OR) and 95% confidence intervals (CI). An assessment was made of the methodological quality of the studies. Moderator analysis and funnel plot analysis were used to explore the sources of between-study heterogeneity. Results: Compared to controls, the odds of work accident was found to be nearly double in workers with OSA (OR = 2.18; 95% CI = 1.53–3.10). Occupational driving was associated with a higher effect size. Conclusions: OSA is an underdiagnosed nonoccupational disease that has a strong adverse effect on work accidents. The nearly twofold increased odds of work accidents in subjects with OSA calls for workplace screening in selected safety-sensitive occupations. Commentary: A commentary on this article appears in this issue on page 1171. Citation: Garbarino S, Guglielmi O, Sanna A, Mancardi GL, Magnavita N. Risk of occupational accidents in workers with obstructive sleep apnea: systematic review and meta-analysis. SLEEP 2016;39(6):1211–1218. PMID:26951401

Analysis of the 1957-1958 Soviet nuclear accident

International Nuclear Information System (INIS)

Trabalka, J.R.; Eyman, L.D.; Auerbach, S.I.

1980-01-01

The presence of an extensive environmental contamination zone in Chelibinsk Province of the Soviet Union, associated with an accident in the winter of 1957 to 1958 involving the atmospheric release of fission wastes, appears to have been confirmed, primarily by an analysis of the Soviet radioecology literature. The contamination zone is estimated to contain 10(5) to 10(6) curies of strontium-90. A plausible explanation for the incident is the use of now-obsolete techniques for waste storage and cesium-137 isotope separation. Radioactive contamination appears to have resulted in resettlement of the human population from a significant area (100 to 1000 square kilometers). It therefore seems imperative to obtain a complete explanation of the cause (or causes) and consequences of the accident; Soviet experience gained in the application of corrective measures would be invaluable to the world nuclear community

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

Analysis of accident caused by temperature increase at the RA reactor in Vinca

International Nuclear Information System (INIS)

Afgan, N; Kulundzic, P.

1964-06-01

The objective of this work was to determine the maximum time interval without accident due to mechanical failures. The following accidents caused by mechanical failures are taken into account: loss of moderator flow, moderator leaking with or without circulation, and accidents caused by removal of fuel channels from the reactor vessel. Obtained numerical and experimental results are presented. Experimental device installed at the reactor was used for verification of calculation results. The analysis was done for the most damaging conditions and thus the obtained results represent the lowest boundary values [sr

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.

Specific features of RBMK severe accidents progression and approach to the accident management

International Nuclear Information System (INIS)

Vasilevskij, V.P.; Nikitin, Yu.M.; Petrov, A.A.; Potapov, A.A.; Cherkashov, Yu.M.

2001-01-01

Fundamental construction features of the LWGR facilities (absence of common external containment shell, disintegrated circulation circuit and multichannel reactor core, positive vapor reactivity coefficient, high mass of thermally capacious graphite moderator) predetermining development of assumed heavy non-projected accidents and handling them are treated. Rating the categories of the reactor core damages for non-projected accidents and accident types producing specific grope of damages is given. Passing standard non-projected accidents, possible methods of attack accident consequences, as well as methods of calculated analysis of non-projected accidents are demonstrated [ru

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.

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.

Systems thinking, the Swiss Cheese Model and accident analysis: a comparative systemic analysis of the Grayrigg train derailment using the ATSB, AcciMap and STAMP models.

Science.gov (United States)

Underwood, Peter; Waterson, Patrick

2014-07-01

The Swiss Cheese Model (SCM) is the most popular accident causation model and is widely used throughout various industries. A debate exists in the research literature over whether the SCM remains a viable tool for accident analysis. Critics of the model suggest that it provides a sequential, oversimplified view of accidents. Conversely, proponents suggest that it embodies the concepts of systems theory, as per the contemporary systemic analysis techniques. The aim of this paper was to consider whether the SCM can provide a systems thinking approach and remain a viable option for accident analysis. To achieve this, the train derailment at Grayrigg was analysed with an SCM-based model (the ATSB accident investigation model) and two systemic accident analysis methods (AcciMap and STAMP). The analysis outputs and usage of the techniques were compared. The findings of the study showed that each model applied the systems thinking approach. However, the ATSB model and AcciMap graphically presented their findings in a more succinct manner, whereas STAMP more clearly embodied the concepts of systems theory. The study suggests that, whilst the selection of an analysis method is subject to trade-offs that practitioners and researchers must make, the SCM remains a viable model for accident analysis. Copyright © 2013 Elsevier Ltd. All rights reserved.

Chemical considerations in severe accident analysis

International Nuclear Information System (INIS)

Malinauskas, A.P.; Kress, T.S.

1988-01-01

The Reactor Safety Study presented the first systematic attempt to include fission product physicochemical effects in the determination of expected consequences of hypothetical nuclear reactor power plant accidents. At the time, however, the data base was sparse, and the treatment of fission product behavior was not entirely consistent or accurate. Considerable research has since been performed to identify and understand chemical phenomena that can occur in the course of a nuclear reactor accident, and how these phenomena affect fission product behavior. In this report, the current status of our understanding of the chemistry of fission products in severe core damage accidents is summarized and contrasted with that of the Reactor Safety Study

Construction industry accidents in Spain.

Science.gov (United States)

Camino López, Miguel A; Ritzel, Dale O; Fontaneda, Ignacio; González Alcantara, Oscar J

2008-01-01

This paper analyzed industrial accidents that take place on construction sites and their severity. Eighteen variables were studied. We analyzed the influence of each of these with respect to the severity and fatality of the accident. This descriptive analysis was grounded in 1,630,452 accidents, representing the total number of accidents suffered by workers in the construction sector in Spain over the period 1990-2000. It was shown that age, type of contract, time of accident, length of service in the company, company size, day of the week, and the remainder of the variables under analysis influenced the seriousness of the accident. IMPACT ON INJURY PREVENTION: The results obtained show that different training was needed, depending on the severity of accidents, for different age, length of service in the company, organization of work, and time when workers work. The research provides an insight to the likely causes of construction injuries in Spain. As a result of the analysis, industries and governmental agencies in Spain can start to provide appropriate strategies and training to the construction workers.

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

Analysis of Sertraline in Postmortem Fluids and Tissues in 11 Aviation Accident Victims

Science.gov (United States)

2012-11-01

likely undergoes significant postmortem redistribution. 17. Key Words 18. Distribution Statement Forensic Toxicology , Sertraline, Norsertraline... Toxicology .. Forensic Sci Int,.142:.75-100.(2004) . 29 .. Skopp,.G ..Postmortem.Toxicology .. Forensic Sci Med Pathol,.6:.314-25.(2010) . ... toxicological . analysis. on. specimens.from.….aircraft.accident.fatalities”.and.“in- vestigate.….general.aviation.and.air.carrier.accidents. and. search

Statistical Analysis And Treatment Of Accident Black Spots: A Case Study Of Nandyal Mandal

Science.gov (United States)

Sudharshan Reddy, B.; Vishnu Vardhan Reddy, L.; Sreenivasa Reddy, G., Dr

2017-08-01

Background: Increased, economic activity raised the consumption levels of the people across the country. This created scope for increase in travel and transportation. The increase in the vehicles since last 10 years has put lot of pressure on the existing roads and ultimately resulting in road accidents. Nandyal Mandal is located in the Kurnool district of Andhra Pradesh and well developed in both agricultural and industrial sectors after Kurnool. 567 accidents occurred in the last seven years at 143 locations shows the severity of the accidents in the Nandyal Mandal. There is a need to carry out some work in the Nandyal Mandal to improve the accidents black spots for reducing the accidents. Methods: Last seven years (2010-2016) of accident data collected from Police Stations. Weighted Severity Index (WSI), a scientific method is used for identifying the accident black spots. Statistical analysis has carried out for the collected data using Chi-Square Test to determine the independence of accidents with other attributes. Chi-Square Goodness of fit test conducted for test whether the accidents are occurring by chance or following any pattern. Results: WSI values are determined for the 143 locations. The Locations with high WSI are treated as accident black spots. Five black spots are taken for field study. After field observations and interaction with the public, some improvements are suggested for improving the accident black spots. There is no relationship between the severity of accidents and the other attributes like month, season, day, hours in day and the age group except type of vehicle. Road accidents are distributed throughout the Year, Month and Season. Road accidents are not distributed throughout the day.

Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-yeop; Lim, Ho-Gon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2016-10-15

Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study.

Sensitivity Analysis of Evacuation Speed in Hypothetical NPP Accident by Earthquake

International Nuclear Information System (INIS)

Kim, Sung-yeop; Lim, Ho-Gon

2016-01-01

Effective emergency response in emergency situation of nuclear power plant (NPP) can make consequences be different therefore it is regarded important when establishing an emergency response plan and assessing the risk of hypothetical NPP accident. Situation of emergency response can be totally changed when NPP accident caused by earthquake or tsunami is considered due to the failure of roads and buildings by the disaster. In this study evacuation speed has been focused among above various factors and reasonable evacuation speed in earthquake scenario has been investigated. Finally, sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Evacuation scenario can be entirely different in the situation of seismic hazard and the sensitivity analysis of evacuation speed in hypothetical NPP accident by earthquake has been performed in this study. Various references were investigated and earthquake evacuation model has been developed considering that evacuees may convert their evacuation method from using a vehicle to walking when they face the difficulty of using a vehicle due to intense traffic jam, failure of buildings and roads, and etc. The population dose within 5 km / 30 km have been found to be increased in earthquake situation due to decreased evacuation speed and become 1.5 - 2 times in the severest earthquake evacuation scenario set up in this study. It is not agreed that using same emergency response model which is used for normal evacuation situations when performing level 3 probabilistic safety assessment for earthquake and tsunami event. Investigation of data and sensitivity analysis for constructing differentiated emergency response model in the event of seismic hazard has been carried out in this study

Accidents - Chernobyl accident; Accidents - accident de Tchernobyl

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

Analysis of loss-of-coolant accidents in pressurized water reactors

International Nuclear Information System (INIS)

Moldaschl, H.

1982-01-01

Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

Accident sequences and causes analysis in a hydrogen production process

Energy Technology Data Exchange (ETDEWEB)

Jae, Moo Sung; Hwang, Seok Won; Kang, Kyong Min; Ryu, Jung Hyun; Kim, Min Soo; Cho, Nam Chul; Jeon, Ho Jun; Jung, Gun Hyo; Han, Kyu Min; Lee, Seng Woo [Hanyang Univ., Seoul (Korea, Republic of)

2006-03-15

Since hydrogen production facility using IS process requires high temperature of nuclear power plant, safety assessment should be performed to guarantee the safety of facility. First of all, accident cases of hydrogen production and utilization has been surveyed. Based on the results, risk factors which can be derived from hydrogen production facility were identified. Besides the correlation between risk factors are schematized using influence diagram. Also initiating events of hydrogen production facility were identified and accident scenario development and quantification were performed. PSA methodology was used for identification of initiating event and master logic diagram was used for selection method of initiating event. Event tree analysis was used for quantification of accident scenario. The sum of all the leakage frequencies is 1.22x10{sup -4} which is similar value (1.0x10{sup -4}) for core damage frequency that International Nuclear Safety Advisory Group of IAEA suggested as a criteria.

Safety analysis of RA reactor operation, I-III, Part II, Accident analysis

International Nuclear Information System (INIS)

Raisic, N.

1963-02-01

This volume covers the analyses of two types of accidents: accidents caused by uncontrolled reactivity increase, and accidents caused by decrease or loss of cooling. First type of accidents, uncontrolled reactivity insertion could occur due to removal of compensation, regulatory or safety rods, or by increase of heavy water level. Removal of irradiated samples from the core could also cause increase of reactivity. Second type of accidents could occur due to interruption of cooling, loss of water in the secondary cooling loop or loss of water in the primary coolant loop

Accident analysis in research reactors

International Nuclear Information System (INIS)

Adorni, M.; Bousbia-salah, A.; D'Auria, F.; Hamidouche, T.

2007-01-01

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

Analysis on relation between safety input and accidents

Institute of Scientific and Technical Information of China (English)

YAO Qing-guo; ZHANG Xue-mu; LI Chun-hui

2007-01-01

The number of safety input directly determines the level of safety, and there exists dialectical and unified relations between safety input and accidents. Based on the field investigation and reliable data, this paper deeply studied the dialectical relationship between safety input and accidents, and acquired the conclusions. The security situation of the coal enterprises was related to the security input rate, being effected little by the security input scale, and build the relationship model between safety input and accidents on this basis, that is the accident model.

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)

Accident sequence quantification with KIRAP

Energy Technology Data Exchange (ETDEWEB)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP`s cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs.

Accident sequence quantification with KIRAP

International Nuclear Information System (INIS)

Kim, Tae Un; Han, Sang Hoon; Kim, Kil You; Yang, Jun Eon; Jeong, Won Dae; Chang, Seung Cheol; Sung, Tae Yong; Kang, Dae Il; Park, Jin Hee; Lee, Yoon Hwan; Hwang, Mi Jeong.

1997-01-01

The tasks of probabilistic safety assessment(PSA) consists of the identification of initiating events, the construction of event tree for each initiating event, construction of fault trees for event tree logics, the analysis of reliability data and finally the accident sequence quantification. In the PSA, the accident sequence quantification is to calculate the core damage frequency, importance analysis and uncertainty analysis. Accident sequence quantification requires to understand the whole model of the PSA because it has to combine all event tree and fault tree models, and requires the excellent computer code because it takes long computation time. Advanced Research Group of Korea Atomic Energy Research Institute(KAERI) has developed PSA workstation KIRAP(Korea Integrated Reliability Analysis Code Package) for the PSA work. This report describes the procedures to perform accident sequence quantification, the method to use KIRAP's cut set generator, and method to perform the accident sequence quantification with KIRAP. (author). 6 refs

[Comparative analysis of the radionuclide composition in fallout after the Chernobyl and the Fukushima accidents].

Science.gov (United States)

Kotenko, K V; Shinkarev, S M; Abramov, Iu V; Granovskaia, E O; Iatsenko, V N; Gavrilin, Iu I; Margulis, U Ia; Garetskaia, O S; Imanaka, T; Khoshi, M

2012-01-01

The nuclear accident occurred at Fukushima Dai-ichi Nuclear Power Plant (NPP) (March 11, 2011) similarly to the accident at the Chernobyl NPP (April 26, 1986) is related to the level 7 of the INES. It is of interest to make an analysis of the radionuclide composition of the fallout following the both accidents. The results of the spectrometric measurements were used in that comparative analysis. Two areas following the Chernobyl accident were considered: (1) the near zone of the fallout - the Belarusian part of the central spot extended up to 60 km around the Chernobyl NPS and (2) the far zone of the fallout--the "Gomel-Mogilev" spot centered 200 km to the north-northeast of the damaged reactor. In the case of Fukushima accident the near zone up to about 60 km considered. The comparative analysis has been done with respect to refractory radionuclides (95Zr, 95Nb, 141Ce, 144Ce), as well as to the intermediate and volatile radionuclides 103Ru, 106Ru, 131I, 134Cs, 137Cs, 140La, 140Ba and the results of such a comparison have been discussed. With respect to exposure to the public the most important radionuclides are 131I and 137Cs. For the both accidents the ratios of 131I/137Cs in the considered soil samples are in the similar ranges: (3-50) for the Chernobyl samples and (5-70) for the Fukushima samples. Similarly to the Chernobyl accident a clear tendency that the ratio of 131I/137Cs in the fallout decreases with the increase of the ground deposition density of 137Cs within the trace related to a radioactive cloud has been identified for the Fukushima accident. It looks like this is a universal tendency for the ratio of 131I/137Cs versus the 137Cs ground deposition density in the fallout along the trace of a radioactive cloud as a result of a heavy accident at the NPP with radionuclides releases into the environment. This tendency is important for an objective reconstruction of 131I fallout based on the results of 137Cs measurements of soil samples carried out at

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.

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.

3-Dimensional Methodology for the Control Rod Ejection Accident Analysis Using UNICORNTM

International Nuclear Information System (INIS)

Jang, Chan-su; Um, Kil-sup; Ahn, Dawk-hwan; Kim, Yo-han; Sung, Chang-kyung; Song, Jae-seung

2006-01-01

The control rod ejection accident has been analyzed with STRIKIN-II code using the point kinetics model coupled with conservative factors to address the three dimensional aspects. This may result in a severe transient with very high fuel enthalpy deposition. KNFC, under the support of KEPRI and KAERI, is developing 3-dimensional methodology for the rod ejection accident analysis using UNICORNTM (Unified Code of RETRAN, TORC and MASTER). For this purpose, 3-dimensional MASTER-TORC codes, which have been combined with the dynamic-link library by KAERI, are used in the transient analysis of the core and RETRAN code is used to estimate the enthalpy deposition in the hot rod

Phenomenological uncertainty analysis of early containment failure at severe accident of nuclear power plant

Energy Technology Data Exchange (ETDEWEB)

Lee, Su Won

2011-02-15

The severe accident has inherently significant uncertainty due to wide range of conditions and performing experiments, validation and practical application are extremely difficult because of its high temperature and pressure. Although internal and external researches were put into practice, the reference used in Korean nuclear plants were foreign data of 1980s and safety analysis as the probabilistic safety assessment has not applied the newest methodology. Also, it is applied to containment pressure formed into point value as results of thermal hydraulic analysis to identify the probability of containment failure in level 2 PSA. In this paper, the uncertainty analysis methods for phenomena of severe accident influencing early containment failure were developed, the uncertainty analysis that apply Korean nuclear plants using the MELCOR code was performed and it is a point of view to present the distribution of containment pressure as a result of uncertainty analysis. Because early containment failure is important factor of Large Early Release Frequency(LERF) that is used as representative criteria of decision-making in nuclear power plants, it was selected in this paper among various modes of containment failure. Important phenomena of early containment failure at severe accident based on previous researches were comprehended and methodology of 7th steps to evaluate uncertainty was developed. The MELCOR input for analysis of the severe accident reflected natural circulation flow was developed and the accident scenario for station black out that was representative initial event of early containment failure was determined. By reviewing the internal model and correlation for MELCOR model relevant important phenomena of early containment failure, the uncertainty factors which could affect on the uncertainty were founded and the major factors were finally identified through the sensitivity analysis. In order to determine total number of MELCOR calculations which can

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)

Occupational Accidents And Preventive Measures

CERN Document Server

Fassnacht, V

2006-01-01

This report presents the 2005 statistics concerning occupational accidents involving members of the CERN personnel and contractors' personnel. It sets out the accident frequency and severity rates and provides a breakdown of accidents by cause and injury. It also contains a summary analysis of the most serious accidents and the associated recommendations.

[Severe parachuting accident. Analysis of 122 cases].

Science.gov (United States)

Krauss, U; Mischkowsky, T

1993-06-01

Based on a population of 122 severely injured patients the causes of paragliding accidents and the patterns of injury are analyzed. A questionnaire is used to establish a sport-specific profile for the paragliding pilot. The lower limbs (55.7%) and the lower parts of the spine (45.9%) are the most frequently injured parts of the body. There is a high risk of multiple injuries after a single accident because of the tremendous axial power. The standard of equipment is good in over 90% of the cases. Insufficient training and failure to take account of geographical and meteorological conditions are the main determinants of accidents sustained by paragliders, most of whom are young. Nevertheless, 80% of our patients want to continue paragliding. Finally some advice is given on how to prevent paragliding accidents and injuries.

Fast Transient And Spatially Non-Homogenous Accident Analysis Of Two-Dimensional Cylindrical Nuclear Reactor

International Nuclear Information System (INIS)

Yulianti, Yanti; Su'ud, Zaki; Waris, Abdul; Khotimah, S. N.; Shafii, M. Ali

2010-01-01

The research about fast transient and spatially non-homogenous nuclear reactor accident analysis of two-dimensional nuclear reactor has been done. This research is about prediction of reactor behavior is during accident. In the present study, space-time diffusion equation is solved by using direct methods which consider spatial factor in detail during nuclear reactor accident simulation. Set of equations that obtained from full implicit finite-difference discretization method is solved by using iterative methods ADI (Alternating Direct Implicit). The indication of accident is decreasing macroscopic absorption cross-section that results large external reactivity. The power reactor has a peak value before reactor has new balance condition. Changing of temperature reactor produce a negative Doppler feedback reactivity. The reactivity will reduce excess positive reactivity. Temperature reactor during accident is still in below fuel melting point which is in secure condition.

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)

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)

First international workshop on severe accidents and their consequences. [Chernobyl Accident

Energy Technology Data Exchange (ETDEWEB)

1989-07-01

An international workshop on past severe nuclear accidents and their consequences was held in Dagomys region of Sochi, USSR on October 30--November 3, 1989. The plan of this meeting was approved by the USSR Academy of Sciences and by the USSR State Committee of the Utilization of Atomic Energy. The meeting was held under the umbrella of the ANS-SNS agreement of cooperation. Topics covered include analysis of the Chernobyl accident, safety measures for RBMK type reactors and consequences of the Chernobyl accident including analysis of the ecological, genetic and psycho-social factors. Separate reports are processed separately for the data bases. (CBS)

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

Science.gov (United States)

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

2011-01-01

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

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

Analysis of reactivity insertion accidents in PWR reactors

International Nuclear Information System (INIS)

Camargo, C.T.M.

1978-06-01

A calculation model to analyze reactivity insertion accidents in a PWR reactor was developed. To analyze the nuclear power transient, the AIREK-III code was used, which simulates the conventional point-kinetic equations with six groups of delayed neutron precursors. Some modifications were made to generalize and to adapt the program to solve the proposed problems. A transient thermal analysis model was developed which simulates the heat transfer process in a cross section of a UO 2 fuel rod with Zircalloy clad, a gap fullfilled with Helium gas and the correspondent coolant channel, using as input the nulcear power transient calculated by AIREK-III. The behavior of ANGRA-i reactor was analized during two types of accidents: - uncontrolled rod withdrawal from subcritical condition; - uncontrolled rod withdrawal at power. The results and conclusions obtained will be used in the license process of the Unit 1 of the Central Nuclear Almirante Alvaro Alberto. (Author) [pt

Uncertainty and sensitivity analysis in nuclear accident consequence assessment

International Nuclear Information System (INIS)

Karlberg, Olof.

1989-01-01

This report contains the results of a four year project in research contracts with the Nordic Cooperation in Nuclear Safety and the National Institute for Radiation Protection. An uncertainty/sensitivity analysis methodology consisting of Latin Hypercube sampling and regression analysis was applied to an accident consequence model. A number of input parameters were selected and the uncertainties related to these parameter were estimated within a Nordic group of experts. Individual doses, collective dose, health effects and their related uncertainties were then calculated for three release scenarios and for a representative sample of meteorological situations. From two of the scenarios the acute phase after an accident were simulated and from one the long time consequences. The most significant parameters were identified. The outer limits of the calculated uncertainty distributions are large and will grow to several order of magnitudes for the low probability consequences. The uncertainty in the expectation values are typical a factor 2-5 (1 Sigma). The variation in the model responses due to the variation of the weather parameters is fairly equal to the parameter uncertainty induced variation. The most important parameters showed out to be different for each pathway of exposure, which could be expected. However, the overall most important parameters are the wet deposition coefficient and the shielding factors. A general discussion of the usefulness of uncertainty analysis in consequence analysis is also given. (au)

Accident proneness, does it exist? A review and meta-analysis

OpenAIRE

Visser, Ellen; Pijl, Ysbrand J.; Stolk, Ronald P.; Neeleman, Jan; Rosmalen, Judith G. M.

2007-01-01

Accident related health problems have been suggested to cluster within persons. This phenomenon became known as accident proneness and has been a subject of many discussions. This study provides an overview of accident proneness. Therefore, 79 articles with empirical data on accident rates were identified from databases Embase, Medline, and Psychinfo. First, definitions of accidents varied highly, but most studies focused on accidents resulting in injuries requiring medical attention. Second,...

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

Pressure Load Analysis during Severe Accidents for the Evaluation of Late Containment Failure in OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Park, S. Y.; Ahn, K. I. [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2014-05-15

The MAAP code is a system level computer code capable of performing integral analyses of potential severe accident progressions in nuclear power plants, whose main purpose is to support a level 2 probabilistic safety assessment or severe accident management strategy developments. The code employs lots of user-options for supporting a sensitivity and uncertainty analysis. The present application is mainly focused on determining an estimate of the containment building pressure load caused by severe accident sequences. Key modeling parameters and phenomenological models employed for the present uncertainty analysis are closely related to in-vessel hydrogen generation, gas combustion in the containment, corium distribution in the containment after a reactor vessel failure, corium coolability in the reactor cavity, and molten-corium interaction with concrete. The phenomenology of severe accidents is extremely complex. In this paper, a sampling-based phenomenological uncertainty analysis was performed to statistically quantify uncertainties associated with the pressure load of a containment building for a late containment failure evaluation, based on the key modeling parameters employed in the MAAP code and random samples for those parameters. Phenomenological issues surrounding the late containment failure mode are highly complex. Included are the pressurization owing to steam generation in the cavity, molten corium-concrete interaction, late hydrogen burn in the containment, and the secondary heat removal availability. The methodology and calculation results can be applied for the optimum assessment of a late containment failure model. The accident sequences considered were a loss of coolant accidents and loss of offsite accidents expected in the OPR-1000 plant. As a result, uncertainties addressed in the pressure load of the containment building were quantified as a function of time. A realistic evaluation of the mean and variance estimates provides a more complete

Radiation accidents

International Nuclear Information System (INIS)

Nenot, J.C.

1996-01-01

Analysis of radiation accidents over a 50 year period shows that simple cases, where the initiating events were immediately recognised, the source identified and under control, the medical input confined to current handling, were exceptional. In many cases, the accidents were only diagnosed when some injuries presented by the victims suggested the radiological nature of the cause. After large-scale accidents, the situation becomes more complicated, either because of management or medical problems, or both. The review of selected accidents which resulted in severe consequences shows that most of them could have been avoided; lack of regulations, contempt for rules, human failure and insufficient training have been identified as frequent initiating parameters. In addition, the situation was worsened because of unpreparedness, insufficient planning, unadapted resources, and underestimation of psychosociological aspects. (author)

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.)

Analysis of selected factors that generate the costs of accidents at work using the Polish construction industry as an example

Directory of Open Access Journals (Sweden)

Hoła Anna

2016-01-01

Full Text Available The paper presents analysis of selected factors that generate the costs of accidents at work using the Polish construction industry as an example. The individual components of the cost of accidents have been identified. Using the statistical data published by the Central Statistical Office, the impact on the size of the cost of accidents at work of such factors as the lost time of an injured person, the lost time of other people involved in the removal of accident effects and also material losses caused by an accident, was analysed. On the basis of the conducted analysis, conclusions regarding economic losses due to accidents were formulated.

Analysis of ASTEC code adaptability to severe accident simulation for CANDU type reactors

International Nuclear Information System (INIS)

Constantin, Marin; Rizoiu, Andrei

2008-01-01

In order to prepare the adaptation of the ASTEC code to CANDU NPP severe accident analysis two kinds of activities were performed: - analyses of the ASTEC modules from the point of view of models and options, followed by CANDU exploratory calculation for the appropriate modules/models; - preparing the specifications for ASTEC adaptation for CANDU NPP. The paper is structured in three parts: - a comparison of PWR and CANDU concepts (from the point of view of severe accident phenomena); - exploratory calculations with some ASTEC modules- SOPHAEROS, CPA, IODE, CESAR, DIVA - for CANDU type reactors specific problems; - development needs analysis - algorithms, methods, modules. (authors)

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)

An association between dietary habits and traffic accidents in patients with chronic liver disease: A data-mining analysis.

Science.gov (United States)

Kawaguchi, Takumi; Suetsugu, Takuro; Ogata, Shyou; Imanaga, Minami; Ishii, Kumiko; Esaki, Nao; Sugimoto, Masako; Otsuyama, Jyuri; Nagamatsu, Ayu; Taniguchi, Eitaro; Itou, Minoru; Oriishi, Tetsuharu; Iwasaki, Shoko; Miura, Hiroko; Torimura, Takuji

2016-05-01

The incidence of traffic accidents in patients with chronic liver disease (CLD) is high in the USA. However, the characteristics of patients, including dietary habits, differ between Japan and the USA. The present study investigated the incidence of traffic accidents in CLD patients and the clinical profiles associated with traffic accidents in Japan using a data-mining analysis. A cross-sectional study was performed and 256 subjects [148 CLD patients (CLD group) and 106 patients with other digestive diseases (disease control group)] were enrolled; 2 patients were excluded. The incidence of traffic accidents was compared between the two groups. Independent factors for traffic accidents were analyzed using logistic regression and decision-tree analyses. The incidence of traffic accidents did not differ between the CLD and disease control groups (8.8 vs. 11.3%). The results of the logistic regression analysis showed that yoghurt consumption was the only independent risk factor for traffic accidents (odds ratio, 0.37; 95% confidence interval, 0.16-0.85; P=0.0197). Similarly, the results of the decision-tree analysis showed that yoghurt consumption was the initial divergence variable. In patients who consumed yoghurt habitually, the incidence of traffic accidents was 6.6%, while that in patients who did not consume yoghurt was 16.0%. CLD was not identified as an independent factor in the logistic regression and decision-tree analyses. In conclusion, the difference in the incidence of traffic accidents in Japan between the CLD and disease control groups was insignificant. Furthermore, yoghurt consumption was an independent negative risk factor for traffic accidents in patients with digestive diseases, including CLD.

An association between dietary habits and traffic accidents in patients with chronic liver disease: A data-mining analysis

Science.gov (United States)

KAWAGUCHI, TAKUMI; SUETSUGU, TAKURO; OGATA, SHYOU; IMANAGA, MINAMI; ISHII, KUMIKO; ESAKI, NAO; SUGIMOTO, MASAKO; OTSUYAMA, JYURI; NAGAMATSU, AYU; TANIGUCHI, EITARO; ITOU, MINORU; ORIISHI, TETSUHARU; IWASAKI, SHOKO; MIURA, HIROKO; TORIMURA, TAKUJI

2016-01-01

The incidence of traffic accidents in patients with chronic liver disease (CLD) is high in the USA. However, the characteristics of patients, including dietary habits, differ between Japan and the USA. The present study investigated the incidence of traffic accidents in CLD patients and the clinical profiles associated with traffic accidents in Japan using a data-mining analysis. A cross-sectional study was performed and 256 subjects [148 CLD patients (CLD group) and 106 patients with other digestive diseases (disease control group)] were enrolled; 2 patients were excluded. The incidence of traffic accidents was compared between the two groups. Independent factors for traffic accidents were analyzed using logistic regression and decision-tree analyses. The incidence of traffic accidents did not differ between the CLD and disease control groups (8.8 vs. 11.3%). The results of the logistic regression analysis showed that yoghurt consumption was the only independent risk factor for traffic accidents (odds ratio, 0.37; 95% confidence interval, 0.16–0.85; P=0.0197). Similarly, the results of the decision-tree analysis showed that yoghurt consumption was the initial divergence variable. In patients who consumed yoghurt habitually, the incidence of traffic accidents was 6.6%, while that in patients who did not consume yoghurt was 16.0%. CLD was not identified as an independent factor in the logistic regression and decision-tree analyses. In conclusion, the difference in the incidence of traffic accidents in Japan between the CLD and disease control groups was insignificant. Furthermore, yoghurt consumption was an independent negative risk factor for traffic accidents in patients with digestive diseases, including CLD. PMID:27123257

Aspects of risk analysis application to estimation of nuclear accidents and tests consequences and intervention management

International Nuclear Information System (INIS)

Demin, V.F.; Hedemann-Jensen, P.; Rolevich, I.V.; Schneider, T.S.; Sobolev, B.G.

1996-01-01

For assessment of accident consequences and a post-accident management a risk analysis methodology and data bank (BARD) with allowance for radiation and non-radiation risk causes should be developed and used. Aspects of these needs and developments are considered. Some illustrative results of health risk estimation made with BARD for the Bryansk region territory with relatively high radioactive contamination from the Chernobyl accident are presented

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

Analysis of Moderator System Failure Accidents by Using New Method for Wolsong-1 CANDU 6 Reactor

Energy Technology Data Exchange (ETDEWEB)

Jin, Dongsik; Kim, Jonghyun; Cho, Cheonhwey [Atomic Creative Technology Co., Ltd., Daejeon (Korea, Republic of); Kim, Sungmin [Korea Hydro and Nuclear Power Co., Ltd., Daejeon (Korea, Republic of)

2013-05-15

To reconfirm the safety of moderator system failure accidents, the safety analysis by using the reactor physics code, RFSP-IST, coupled with the thermal hydraulics code, CATHENA is performed additionally. In the present paper, the newly developed analysis method is briefly described and the results obtained from the moderator system failure accident simulations for Wolsong-1 CANDU 6 reactor by using the new method are summarized. The safety analysis of the moderator system failure accidents for Wolsong-1 CANDU 6 reactor was carried out by using the new code system, i. e., CATHENA and RFSP-IST, instead of the non-IST old codes, namely, SMOKIN G-2 and MODSTBOIL. The analysis results by using the new method revealed as same with the results by using the old method that the fuel integrity is warranted because the localized power peak remained well below the limits and, most importantly, the reactor operation enters into the self-shutdown mode due to the substantial loss of moderator D{sub 2}O inventory from the moderator system. In the analysis results obtained by using the old method, it was predicted that the ROP trip conditions occurred for the transient cases which are also studied in the present paper. But, in the new method, it was found that the ROP trip conditions did not occur. Consequently, in the safety analysis performed additionally by using the new method, the safety of moderator system failure accidents was reassured. In the future, the new analysis method by using the IST codes instead of the non-IST old codes for the moderator system failure accidents is strongly recommended.

Analysis of Occupational Accidents in Underground and Surface Mining in Spain Using Data-Mining Techniques.

Science.gov (United States)

Sanmiquel, Lluís; Bascompta, Marc; Rossell, Josep M; Anticoi, Hernán Francisco; Guash, Eduard

2018-03-07

An analysis of occupational accidents in the mining sector was conducted using the data from the Spanish Ministry of Employment and Social Safety between 2005 and 2015, and data-mining techniques were applied. Data was processed with the software Weka. Two scenarios were chosen from the accidents database: surface and underground mining. The most important variables involved in occupational accidents and their association rules were determined. These rules are composed of several predictor variables that cause accidents, defining its characteristics and context. This study exposes the 20 most important association rules in the sector-either surface or underground mining-based on the statistical confidence levels of each rule as obtained by Weka. The outcomes display the most typical immediate causes, along with the percentage of accidents with a basis in each association rule. The most important immediate cause is body movement with physical effort or overexertion, and the type of accident is physical effort or overexertion. On the other hand, the second most important immediate cause and type of accident are different between the two scenarios. Data-mining techniques were chosen as a useful tool to find out the root cause of the accidents.

Analysis of Occupational Accidents in Underground and Surface Mining in Spain Using Data-Mining Techniques

Science.gov (United States)

Sanmiquel, Lluís; Bascompta, Marc; Rossell, Josep M.; Anticoi, Hernán Francisco; Guash, Eduard

2018-01-01

An analysis of occupational accidents in the mining sector was conducted using the data from the Spanish Ministry of Employment and Social Safety between 2005 and 2015, and data-mining techniques were applied. Data was processed with the software Weka. Two scenarios were chosen from the accidents database: surface and underground mining. The most important variables involved in occupational accidents and their association rules were determined. These rules are composed of several predictor variables that cause accidents, defining its characteristics and context. This study exposes the 20 most important association rules in the sector—either surface or underground mining—based on the statistical confidence levels of each rule as obtained by Weka. The outcomes display the most typical immediate causes, along with the percentage of accidents with a basis in each association rule. The most important immediate cause is body movement with physical effort or overexertion, and the type of accident is physical effort or overexertion. On the other hand, the second most important immediate cause and type of accident are different between the two scenarios. Data-mining techniques were chosen as a useful tool to find out the root cause of the accidents. PMID:29518921

Analysis of Occupational Accidents in Underground and Surface Mining in Spain Using Data-Mining Techniques

Directory of Open Access Journals (Sweden)

Lluís Sanmiquel

2018-03-01

Full Text Available An analysis of occupational accidents in the mining sector was conducted using the data from the Spanish Ministry of Employment and Social Safety between 2005 and 2015, and data-mining techniques were applied. Data was processed with the software Weka. Two scenarios were chosen from the accidents database: surface and underground mining. The most important variables involved in occupational accidents and their association rules were determined. These rules are composed of several predictor variables that cause accidents, defining its characteristics and context. This study exposes the 20 most important association rules in the sector—either surface or underground mining—based on the statistical confidence levels of each rule as obtained by Weka. The outcomes display the most typical immediate causes, along with the percentage of accidents with a basis in each association rule. The most important immediate cause is body movement with physical effort or overexertion, and the type of accident is physical effort or overexertion. On the other hand, the second most important immediate cause and type of accident are different between the two scenarios. Data-mining techniques were chosen as a useful tool to find out the root cause of the accidents.

A strategy to the development of a human error analysis method for accident management in nuclear power plants using industrial accident dynamics

International Nuclear Information System (INIS)

Lee, Yong Hee; Kim, Jae Whan; Jung, Won Dae; Ha, Jae Ju

1998-06-01

This technical report describes the early progress of he establishment of a human error analysis method as a part of a human reliability analysis(HRA) method for the assessment of the human error potential in a given accident management strategy. At first, we review the shortages and limitations of the existing HRA methods through an example application. In order to enhance the bias to the quantitative aspect of the HRA method, we focused to the qualitative aspect, i.e., human error analysis(HEA), during the proposition of a strategy to the new method. For the establishment of a new HEA method, we discuss the basic theories and approaches to the human error in industry, and propose three basic requirements that should be maintained as pre-requisites for HEA method in practice. Finally, we test IAD(Industrial Accident Dynamics) which has been widely utilized in industrial fields, in order to know whether IAD can be so easily modified and extended to the nuclear power plant applications. We try to apply IAD to the same example case and develop new taxonomy of the performance shaping factors in accident management and their influence matrix, which could enhance the IAD method as an HEA method. (author). 33 refs., 17 tabs., 20 figs

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.

Explorative spatial analysis of traffic accident statistics and road mortality among the provinces of Turkey.

Science.gov (United States)

Erdogan, Saffet

2009-10-01

The aim of the study is to describe the inter-province differences in traffic accidents and mortality on roads of Turkey. Two different risk indicators were used to evaluate the road safety performance of the provinces in Turkey. These indicators are the ratios between the number of persons killed in road traffic accidents (1) and the number of accidents (2) (nominators) and their exposure to traffic risk (denominator). Population and the number of registered motor vehicles in the provinces were used as denominators individually. Spatial analyses were performed to the mean annual rate of deaths and to the number of fatal accidents that were calculated for the period of 2001-2006. Empirical Bayes smoothing was used to remove background noise from the raw death and accident rates because of the sparsely populated provinces and small number of accident and death rates of provinces. Global and local spatial autocorrelation analyses were performed to show whether the provinces with high rates of deaths-accidents show clustering or are located closer by chance. The spatial distribution of provinces with high rates of deaths and accidents was nonrandom and detected as clustered with significance of Paccidents and deaths were located in the provinces that contain the roads connecting the Istanbul, Ankara, and Antalya provinces. Accident and death rates were also modeled with some independent variables such as number of motor vehicles, length of roads, and so forth using geographically weighted regression analysis with forward step-wise elimination. The level of statistical significance was taken as Paccidents according to denominators in the provinces. The geographically weighted regression analyses did significantly better predictions for both accident rates and death rates than did ordinary least regressions, as indicated by adjusted R(2) values. Geographically weighted regression provided values of 0.89-0.99 adjusted R(2) for death and accident rates, compared with 0

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.

Analysis of Roadway Traffic Accidents Based on Rough Sets and Bayesian Networks

Directory of Open Access Journals (Sweden)

Xiaoxia Xiong

2018-02-01

Full Text Available The paper integrates Rough Sets (RS and Bayesian Networks (BN for roadway traffic accident analysis. RS reduction of attributes is first employed to generate the key set of attributes affecting accident outcomes, which are then fed into a BN structure as nodes for BN construction and accident outcome classification. Such RS-based BN framework combines the advantages of RS in knowledge reduction capability and BN in describing interrelationships among different attributes. The framework is demonstrated using the 100-car naturalistic driving data from Virginia Tech Transportation Institute to predict accident type. Comparative evaluation with the baseline BNs shows the RS-based BNs generally have a higher prediction accuracy and lower network complexity while with comparable prediction coverage and receiver operating characteristic curve area, proving that the proposed RS-based BN overall outperforms the BNs with/without traditional feature selection approaches. The proposed RS-based BN indicates the most significant attributes that affect accident types include pre-crash manoeuvre, driver’s attention from forward roadway to centre mirror, number of secondary tasks undertaken, traffic density, and relation to junction, most of which feature pre-crash driver states and driver behaviours that have not been extensively researched in literature, and could give further insight into the nature of traffic accidents.

Accidents in the construction industry in the Netherlands: An analysis of accident reports using Storybuilder

International Nuclear Information System (INIS)

Ale, B.J.M.; Bellamy, L.J.; Baksteen, H.; Damen, M.; Goossens, L.H.J.; Hale, A.R.; Mud, M.; Oh, J.; Papazoglou, I.A.; Whiston, J.Y.

2008-01-01

As part of an ongoing effort by the Ministry of Social Affairs and Employment of the Netherlands, a research project is being undertaken to construct a causal model for occupational risk. This model should provide quantitative insight into the causes and consequences of occupational accidents. One of the components of the model is a tool to systematically classify and analyse reports of past accidents. This tool 'Storybuilder' was described in earlier papers. In this paper, Storybuilder is used to analyse the causes of accidents reported in the database of the Dutch Labour Inspectorate involving people working in the construction industry. Conclusions are drawn on measures to reduce the accident probability. Some of these conclusions are contrary to common beliefs in the industry

MCC-15: waste/canister accident testing and analysis method

International Nuclear Information System (INIS)

Slate, S.C.; Pulsipher, B.A.; Scott, P.A.

1985-02-01

The Materials Characterization Center (MCC) at the Pacific Northwest Laboratory (PNL) is developing standard tests to characterize the performance of nuclear waste forms under normal and accident conditions. As part of this effort, the MCC is developing MCC-15, Waste/Canister Accident Testing and Analysis. MCC-15 is used to test canisters containing simulated waste forms to provide data on the effects of accidental impacts on the waste form particle size and on canister integrity. The data is used to support the design of transportation and handling equipment and to demonstrate compliance with repository waste acceptance specifications. This paper reviews the requirements that led to the development of MCC-15, describes the test method itself, and presents some early results from tests on canisters representative of those proposed for the Defense Waste Processing Facility (DWPF). 13 references, 6 figures

Accident patterns for construction-related workers: a cluster analysis

Science.gov (United States)

Liao, Chia-Wen; Tyan, Yaw-Yauan

2012-01-01

The construction industry has been identified as one of the most hazardous industries. The risk of constructionrelated workers is far greater than that in a manufacturing based industry. However, some steps can be taken to reduce worker risk through effective injury prevention strategies. In this article, k-means clustering methodology is employed in specifying the factors related to different worker types and in identifying the patterns of industrial occupational accidents. Accident reports during the period 1998 to 2008 are extracted from case reports of the Northern Region Inspection Office of the Council of Labor Affairs of Taiwan. The results show that the cluster analysis can indicate some patterns of occupational injuries in the construction industry. Inspection plans should be proposed according to the type of construction-related workers. The findings provide a direction for more effective inspection strategies and injury prevention programs.

Analysis of Radiation Accident of Non-destructive Inspection and Rational Preparing Bills

International Nuclear Information System (INIS)

Bae, Junwoo; Yoo, Donghan; Kim, Hee Reyoung

2013-01-01

After 2006, according to enactment of Non-destructive Inspection Promotion Act, the number of non-destructive inspection companies and corresponding accident is increased sharply. In this research, it includes characteristic analysis of field of the non-destructive inspection. And from the result of analysis, the purpose of this research is discovering reason for 'Why there is higher accident ratio in non-destructive inspection field, relatively' and preparing effective bill for reducing radiation accidents. The number of worker for non-destructive inspect is increased steadily and non-destructive inspect worker take highest dose. Corresponding to these, it must be needed to prepare bills to protect non-destructive inspect workers. By analysis of accident case, there are many case of carelessness that tools are too heavy to carry it everywhere workers go. And there are some cases caused by deficiency of education that less understanding of radiation and poor operation by less understanding of structure of tools. Also, there is no data specialized to non-destructive inspect field. So, it has to take information from statistical data. Because of this, it is hard to analyze nondestructive inspect field accurately. So, it is required to; preparing rational bills to protect non-destructive inspect workers nondestructive inspect instrument lightening and easy manual which can understandable for low education background people accurate survey data from real worker. To accomplish these, we needs to do; analyze and comprehend the present law about non-destructive inspect worker understand non-destructive inspect instruments accurately and conduct research for developing material developing rational survey to measuring real condition for non-destructive inspect workers

Approaches to accident analysis in recent US Department of Energy environmental impact statements

International Nuclear Information System (INIS)

Mueller, C.; Folga, S.; Nabelssi, B.

1996-01-01

A review of accident analyses in recent US Department of Energy (DOE) Environmental Impact Statements (EISs) was conducted to evaluate the consistency among approaches and to compare these approaches with existing DOE guidance. The review considered several components of an accident analysis: the overall scope, which in turn should reflect the scope of the EIS; the spectrum of accidents considered; the methods and assumptions used to determine frequencies or frequency ranges for the accident sequences; and the assumption and technical bases for developing radiological and chemical atmospheric source terms and for calculating the consequences of airborne releases. The review also considered the range of results generated with respect to impacts on various worker and general populations. In this paper, the findings of these reviews are presented and methods recommended for improving consistency among EISs and bringing them more into line with existing DOE guidance

Analysis of space-time core dynamics on reactor accident at Chernobyl

International Nuclear Information System (INIS)

Takano, Makoto; Shindo, Ryuichi; Yamashita, Kiyonobu; Sawa, Kazuhiro

1987-05-01

Regarding reactor accident at Chernobyl in USSR, core dynamics has been analyzed by COMIC code which solves space-time dependent diffusion equation in three-dimension taking spatial thermohydraulic effect into account. The code was originally developed for high temperature gas-cooled reactors (HTGR), however, has been modified to include light water as coolant, instead of helium, for analysis of the accident. In the analysis, emphasis is placed on spatial effects on core dynamics. The analyses are performed for the cases of modeling the core fully and partially where 6 fuel channels surround one control rod channel. The result shows that the speed of applying void reactivity averaged over the core depends on the power and coolant flow distributions. Therefore, these distributions have potential to influence on the value and the time of peak power estimated by calculation. (author)

Aspects of using a best-estimate approach for VVER safety analysis in reactivity initiated accidents

Energy Technology Data Exchange (ETDEWEB)

Ovdiienko, Iurii; Bilodid, Yevgen; Ieremenko, Maksym [State Scientific and Technical Centre on Nuclear and Radiation, Safety (SSTC N and RS), Kyiv (Ukraine); Loetsch, Thomas [TUEV SUED Industrie Service GmbH, Energie und Systeme, Muenchen (Germany)

2016-09-15

At present time, Ukraine faces the problem of small margins of acceptance criteria in connection with the implementation of a conservative approach for safety evaluations. The problem is particularly topical conducting feasibility analysis of power up-rating for Ukrainian nuclear power plants. Such situation requires the implementation of a best-estimate approach on the basis of an uncertainty analysis. For some kind of accidents, such as loss-of-coolant accident (LOCA), the best estimate approach is, more or less, developed and established. However, for reactivity initiated accident (RIA) analysis an application of best estimate method could be problematical. A regulatory document in Ukraine defines a nomenclature of neutronics calculations and so called ''generic safety parameters'' which should be used as boundary conditions for all VVER-1000 (V-320) reactors in RIA analysis. In this paper the ideas of uncertainty evaluations of generic safety parameters in RIA analysis in connection with the use of the 3D neutron kinetic code DYN3D and the GRS SUSA approach are presented.

Analysis of Fukushima unit 2 accident considering the operating conditions of RCIC system

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung Il, E-mail: sikim@kaeri.re.kr; Park, Jong Hwa; Ha, Kwang Soon; Cho, Song-Won; Song, JinHo

2016-03-15

Highlights: • Fukushima unit 2 accident was analyzed using MELCOR 1.8.6. • RCIC operating conditions were assumed and best case was selected. • Effect of RCIC operating condition on accident scenario was found. - Abstract: A severe accident in Fukushima occurred on March 11, 2011 and units 1, 2 and 3 were damaged severely. A tsunami following an earthquake made the supply of electricity power stop, and the safety systems, which use AC or DC power in plants could not operate properly. It is supposed that the degree of core degradation of unit 2 is less serious than in the other plants, and it was estimated that the operation of reactor core isolation cooling (RCIC) system at the initial stage of the accident minimized the core damage through decay heat removal. Although the operating conditions of the RCIC system are not known clearly, it can be important to analyze the accident scenario of unit 2. In this study, best case of the Fukushima unit 2 accident was presented considering the operating conditions of the RCIC system. The effects of operating condition on core degradation and fission product release rate to environment were also examined. In addition, importance of torus room flooding level in the accident analysis was discussed. MELCOR 1.8.6 was used in this research, and the geometries of plant and operating conditions of safety system were obtained from TEPCO through OECD/NEA BSAF Project.

Accident Analysis Methods and Models — a Systematic Literature Review

NARCIS (Netherlands)

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

2017-01-01

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

Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery.

Science.gov (United States)

Dźwiarek, Marek; Latała, Agata

2016-01-01

This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005-2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc.

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 proneness, does it exist? A review and meta-analysis

NARCIS (Netherlands)

Visser, Ellen; Pijl, Ysbrand J.; Stolk, Ronald P.; Neeleman, Jan; Rosmalen, Judith G. M.

Accident related health problems have been suggested to cluster within persons. This phenomenon became known as accident proneness and has been a subject of many discussions. This study provides an overview of accident proneness. Therefore, 79 articles with empirical data on accident rates were

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 Importance of Bloodstain Pattern Analysis in the Investigation of Road Traffic Accidents: A Case Report

Directory of Open Access Journals (Sweden)

Younis M. Albalooshi

2015-12-01

Full Text Available Bloodstain pattern analysis has become a field of specialization in Forensic sciences and plays an important role in the reconstruction of events at a crime scene. Research, books, and articles have been published on the analysis and interpretation of bloodstain patterns We present a case study of a road traffic accident in which bloodstain pattern analysis helped us to solve the discrepancy between reports produced by forensic examiners and by the forensic biology department. The case was of a 22-year-old man who died immediately and a 31- year-old woman who survived a road traffic accident. They were both found outside their overturned car and it was impossible to ascertain from initial observations which of the victims was driving the car at the time of the accident. An external examination of the man revealed multiple injuries, and the cause of his death was severe brain injury. The woman survived with a fracture of the forearm, dislocated clavicle bone, and other minor injuries. After initial examination of the car and based on the pattern of injuries the deceased received, forensic examiner concluded that the man was the driving the car at the time of accident. On the other hand, the forensic DNA analysis of bloodstains obtained from the driver's seat matched that of the woman, suggesting that she was the driver. This apparent discrepancy directed the forensic examiner to carry out a bloodstain pattern analysis on the driver's seat. The bloodstain pattern analysis helped resolve the discrepancy and enabled the investigators to identify the driver correctly. This case report emphasizes the importance of bloodstain pattern analysis in the reconstruction of cases involving road traffic accidents.

Impact of spatial kinetics in severe accident analysis for a large HWR

International Nuclear Information System (INIS)

Morris, E.E.

1994-01-01

The impact on spatial kinetics on the analysis of severe accidents initiated by the unprotected withdrawal of one or more control rods is investigated for a large heavy water reactor. Large inter- and intra-assembly power shifts are observed, and the importance of detailed geometrical modeling of fuel assemblies is demonstrated. Neglect of space-time effects is shown to lead to erroneous estimates of safety margins, and of accident consequences in the event safety margins are exceeded. The results and conclusions are typical of what would be expected for any large, loosely coupled core

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.

Analysis of accidents at the LPR (Radiochemical Processes Laboratory)

International Nuclear Information System (INIS)

Kaufmann, F.; Boutet, L.I.

1987-01-01

Accidents are defined as not planned events that may result in the emission of significative quantities of radioactive materials to the environment. The pilot plant has been specifically designed to prevent this type of accidents but there still exists the possibility that one or more accidents can be produced during the plant life. In a first phase, the emission of radionuclides to the environment were evaluated for 13 credible accidents. In a second phase, by means of the calculation program SEDA, specially adapted to this purpose, the critical doses of critical group were calculated for each accident. Due to the small capacity of the pilot plant and the long cooling period of treated fuel, it is concluded that the radiological consequences for the external environment are of very small magnitude. In this way, without need of developing complex fault- or event-trees, it is shown that any of the accidents falls into the non acceptable zone of Farmer diagram. (Author)

Analysis of Fukushima Daiichi Nuclear Power Station severe accident using MAAP4.05

International Nuclear Information System (INIS)

Yoo, Jae S.; Suh, Kune Y.; Kim, Dong M.

2011-01-01

Rather extensive reactor core meltdown and partial melt-through that took place at the Fukushima Daiichi nuclear power plants (NPPs) on March 11, 2011 had been caused by a massive earthquake followed by tsunami rarely seen in history. The happening had turned into an unprecedented serious accident since the Chernobyl Unit 4 in 1986 that extended over multiple reactors simultaneously. A previous documentary survey, NUREG-1150 report provides significant insights into how a severe accident might develop at a boiling water reactor (BWR) and the range of consequences. NUREG-1150 did identify the importance of loss of power accidents for a BWR. This paper describes recent analyses of the Fukushima Daiichi NPPs severe accident. Calculations were performed with the MAAP4.05 code by modifying the parameter file for the Peach Bottom Unit 2, a BWR 4 type in the Mark-I containment. Generally this is the same type of reactor as Fukushima Daiichi Units 2 and 3. This resulted in good understanding of the response of this type of early BWRs to prolonged loss of diesel generators and batteries. There is clearly vulnerability with this early type of BWRs which would be less onerous than later existing plant and new build designs. This analysis, however, was based on rather limited amount of information obtained at the time of preparing this report and adopted various estimates and assumptions for conditions necessary to run the analysis. Hence there persists considerable uncertainty in the results. MAAP4.05 was run pursuant to the observed data and chronology of Fukushima Daiichi Units 1 through 3 reported by the Tokyo Electric Power Company (TEPCO) as well as the Japanese Government. Severe accident scenarios have not only gone far beyond the design basis, but also exceeded the extent of multiple breakdowns assumed in the preparation for such accident management measures as the malfunction or loss of all the emergency core cooling system (ECCS) combined with the extended loss of

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)

[Retrospective analysis of 44 childhood drowning accidents].

Science.gov (United States)

Brüning, Caroline; Siekmeyer, Werner; Siekmeyer, Manuela; Merkenschlager, Andreas; Kiess, Wieland

2010-07-01

Worldwide, drowning is the second leading cause of unintentional death and the leading cause of cardiovascular failure for children [1-3]. The number of near-drownings, where the incident is survived for at least 24 hours, is assumed to be four times as high [5]. In the years 1994 until 2008 there were 44 cases of drowning treated at the children's department of the University of Leipzig. This number shows that even in a medical centre drowning incidents are only occasional incidents. Therefore it is important to know the sequelae and handlings to be able to react in case of an emergency. A total of 44 children suffering a drowning accident within the last 48 hours who were treated during the period of 01.01.1994 through 30.06.2008 at the Children's Centre at the University of Leipzig. A retrospective analysis using a structured questionnaire was done. Social demographic data, accident progress, clinical results and progress as well as outcome of the cases were investigated. During the analysed period in the median three children were treated each year after drowning incidents. Clustering in the summer and winter months and on the weekends was recognizable. The median age was 3.33 years and the group of high risk were children aged 1-3 years, especially boys. Sixty percent of the children came from stable social backgrounds. Half of the children suffered from drowning in created swimming pools or ponds, the rest in natural waters, public pools and sources of water in the household. The median submersion lasted 2 minutes. Correlation of submersions below 1 minute with a good, and submersions above 10 minutes with a negative outcome was shown. A Glasgow Coma Scale (GCS) of 3 points (n = 15) and pupils without light reaction (n = 14) were associated with a lethal outcome or residual neurological deficits. Looking at the laboratory values, correlation between severe acidotic pH-values with a very low base excess, high blood sugar as well as high lactate values and a

Severe accident analysis in a two-loop PWR nuclear power plant with the ASTEC code

International Nuclear Information System (INIS)

Sadek, Sinisa; Amizic, Milan; Grgic, Davor

2013-01-01

The ASTEC/V2.0 computer code was used to simulate a hypothetical severe accident sequence in the nuclear power plant Krsko, a 2-loop pressurized water reactor (PWR) plant. ASTEC is an integral code jointly developed by Institut de Radioprotection et de Surete Nucleaire (IRSN, France) and Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS, Germany) to assess nuclear power plant behaviour during a severe accident. The analysis was conducted in 2 steps. First, the steady state calculation was performed in order to confirm the applicability of the plant model and to obtain correct initial conditions for the accident analysis. The second step was the calculation of the station blackout accident with a leakage of the primary coolant through degraded reactor coolant pump seals, which was a small LOCA without makeup capability. Two scenarios were analyzed: one with and one without the auxiliary feedwater (AFW). The latter scenario, without the AFW, resulted in earlier core damage. In both cases, the accident ended with a core melt and a reactor pressure vessel failure with significant release of hydrogen. In addition, results of the ASTEC calculation were compared with results of the RELAP5/SCDAPSIM calculation for the same transient scenario. The results comparison showed a good agreement between predictions of those 2 codes. (orig.)

Professional experience and traffic accidents/near-miss accidents among truck drivers.

Science.gov (United States)

Girotto, Edmarlon; Andrade, Selma Maffei de; González, Alberto Durán; Mesas, Arthur Eumann

2016-10-01

To investigate the relationship between the time working as a truck driver and the report of involvement in traffic accidents or near-miss accidents. A cross-sectional study was performed with truck drivers transporting products from the Brazilian grain harvest to the Port of Paranaguá, Paraná, Brazil. The drivers were interviewed regarding sociodemographic characteristics, working conditions, behavior in traffic and involvement in accidents or near-miss accidents in the previous 12 months. Subsequently, the participants answered a self-applied questionnaire on substance use. The time of professional experience as drivers was categorized in tertiles. Statistical analyses were performed through the construction of models adjusted by multinomial regression to assess the relationship between the length of experience as a truck driver and the involvement in accidents or near-miss accidents. This study included 665 male drivers with an average age of 42.2 (±11.1) years. Among them, 7.2% and 41.7% of the drivers reported involvement in accidents and near-miss accidents, respectively. In fully adjusted analysis, the 3rd tertile of professional experience (>22years) was shown to be inversely associated with involvement in accidents (odds ratio [OR] 0.29; 95% confidence interval [CI] 0.16-0.52) and near-miss accidents (OR 0.17; 95% CI 0.05-0.53). The 2nd tertile of professional experience (11-22 years) was inversely associated with involvement in accidents (OR 0.63; 95% CI 0.40-0.98). An evident relationship was observed between longer professional experience and a reduction in reporting involvement in accidents and near-miss accidents, regardless of age, substance use, working conditions and behavior in traffic. Copyright © 2016 Elsevier Ltd. All rights reserved.

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)

Radiation accidents

International Nuclear Information System (INIS)

Poplavskij, K.K.; Smorodintseva, G.I.

1978-01-01

On the basis of a critical analysis of the available data on causes and consequences of radiation accidents (RA), a classification of RA by severity (five groups of accidents) according to biomedical consequences and categories of exposed personnel is proposed. A RA is defined and its main characteristics are described. Methods of RA prevention are proposed, as is a plan of specific measures to deal with RA in accordance with the proposed classification

Accident selection methodology for TA-55 FSAR

International Nuclear Information System (INIS)

Letellier, B.C.; Pan, P.Y.; Sasser, M.K.

1995-01-01

In the past, the selection of representative accidents for refined analysis from the numerous scenarios identified in hazards analyses (HAs) has involved significant judgment and has been difficult to defend. As part of upgrading the Final Safety Analysis Report (FSAR) for the TA-55 plutonium facility at the Los Alamos National Laboratory, an accident selection process was developed that is mostly mechanical and reproducible in nature and fulfills the requirements of the Department of Energy (DOE) Standard 3009 and DOE Order 5480.23. Among the objectives specified by this guidance are the requirements that accident screening (1) consider accidents during normal and abnormal operating conditions, (2) consider both design basis and beyond design basis accidents, (3) characterize accidents by category (operational, natural phenomena, etc.) and by type (spill, explosion, fire, etc.), and (4) identify accidents that bound all foreseeable accident types. The accident selection process described here in the context of the TA-55 FSAR is applicable to all types of DOE facilities

Analysis of Construction Accidents in Turkey and Responsible Parties

Science.gov (United States)

GÜRCANLI, G. Emre; MÜNGEN, Uğur

2013-01-01

Construction is one of the world’s biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972–2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00–12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases. PMID:24077446

Analysis of construction accidents in Turkey and responsible parties.

Science.gov (United States)

Gürcanli, G Emre; Müngen, Uğur

2013-01-01

Construction is one of the world's biggest industry that includes jobs as diverse as building, civil engineering, demolition, renovation, repair and maintenance. Construction workers are exposed to a wide variety of hazards. This study analyzes 1,117 expert witness reports which were submitted to criminal and labour courts. These reports are from all regions of the country and cover the period 1972-2008. Accidents were classified by the consequence of the incident, time and main causes of the accident, construction type, occupation of the victim, activity at time of the accident and party responsible for the accident. Falls (54.1%), struck by thrown/falling object (12.9%), structural collapses (9.9%) and electrocutions (7.5%) rank first four places. The accidents were most likely between the hours 15:00 and 17:00 (22.6%), 10:00-12:00 (18.7%) and just after the lunchtime (9.9%). Additionally, the most common accidents were further divided into sub-types. Expert-witness assessments were used to identify the parties at fault and what acts of negligence typically lead to accidents. Nearly two thirds of the faulty and negligent acts are carried out by the employers and employees are responsible for almost one third of all cases.

Analysis of occupational accidents: prevention through the use of additional technical safety measures for machinery

Science.gov (United States)

Dźwiarek, Marek; Latała, Agata

2016-01-01

This article presents an analysis of results of 1035 serious and 341 minor accidents recorded by Poland's National Labour Inspectorate (PIP) in 2005–2011, in view of their prevention by means of additional safety measures applied by machinery users. Since the analysis aimed at formulating principles for the application of technical safety measures, the analysed accidents should bear additional attributes: the type of machine operation, technical safety measures and the type of events causing injuries. The analysis proved that the executed tasks and injury-causing events were closely connected and there was a relation between casualty events and technical safety measures. In the case of tasks consisting of manual feeding and collecting materials, the injuries usually occur because of the rotating motion of tools or crushing due to a closing motion. Numerous accidents also happened in the course of supporting actions, like removing pollutants, correcting material position, cleaning, etc. PMID:26652689

Light water reactor severe accident seminar. Seminar presentation manual

International Nuclear Information System (INIS)

2004-01-01

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

Light water reactor severe accident seminar. Seminar presentation manual

Energy Technology Data Exchange (ETDEWEB)

NONE

2004-07-01

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

PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

International Nuclear Information System (INIS)

O'Kula, K.R.; Thoman, D.C.; Lowrie, J.; Keller, A.

2008-01-01

Department of Energy (DOE) accident analysis for establishing the required control sets for nuclear facility safety applies a series of simplifying, reasonably conservative assumptions regarding inputs and methodologies for quantifying dose consequences. Most of the analytical practices are conservative, have a technical basis, and are based on regulatory precedent. However, others are judgmental and based on older understanding of phenomenology. The latter type of practices can be found in modeling hypothetical releases into the atmosphere and the subsequent exposure. Often the judgments applied are not based on current technical understanding but on work that has been superseded. The objective of this paper is to review the technical basis for the major inputs and assumptions in the quantification of consequence estimates supporting DOE accident analysis, and to identify those that could be reassessed in light of current understanding of atmospheric dispersion and radiological exposure. Inputs and assumptions of interest include: Meteorological data basis; Breathing rate; and Inhalation dose conversion factor. A simple dose calculation is provided to show the relative difference achieved by improving the technical bases

Determinants of traffic accident mortality in The Netherlands: a geographical analysis

NARCIS (Netherlands)

van Beeck, E. F.; Mackenbach, J. P.; Looman, C. W.; Kunst, A. E.

1991-01-01

In the Netherlands, a country with one of the lowest levels of traffic accident mortality in the world, large regional mortality differences can be observed. An analysis was performed of the contribution of regional differences in traffic mobility (kilometers travelled/person-years), injury rate

An analysis of postulated accident for 49-2 Swimming Pool Reactor

International Nuclear Information System (INIS)

Wang Yongqing; Cu Shaochu; Wang Liugui; Zhang Zengqing

1990-01-01

The thermal hydrodynamic code RETRAN-02 is used for safety analysis of Swimming Pool Reactor. Accident of partial-loss of flow, loss of offsite electric power and unexpected reactivity insertion are analysed and discussed. These results will be helpful for operation safety of the reactor

Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines

Directory of Open Access Journals (Sweden)

Aikaterini D. Baka

2014-09-01

Full Text Available Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury that occurred because the power supply was reconnected inadvertently or by mistake.

Analysis of Two Electrocution Accidents in Greece that Occurred due to Unexpected Re-energization of Power Lines.

Science.gov (United States)

Baka, Aikaterini D; Uzunoglu, Nikolaos K

2014-09-01

Investigation and analysis of accidents are critical elements of safety management. The over-riding purpose of an organization in carrying out an accident investigation is to prevent similar accidents, as well as seek a general improvement in the management of health and safety. Hundreds of workers have suffered injuries while installing, maintaining, or servicing machinery and equipment due to sudden re-energization of power lines. This study presents and analyzes two electrical accidents (1 fatal injury and 1 serious injury) that occurred because the power supply was reconnected inadvertently or by mistake.

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)

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1997-11-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously. Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence

Extension of ship accident analysis to multiple-package shipments

International Nuclear Information System (INIS)

Mills, G.S.; Neuhauser, K.S.

1998-01-01

Severe ship accidents and the probability of radioactive material release from spent reactor fuel casks were investigated previously (Spring, 1995). Other forms of RAM, e.g., plutonium oxide powder, may be shipped in large numbers of packagings rather than in one to a few casks. These smaller, more numerous packagings are typically placed in ISO containers for ease of handling, and several ISO containers may be placed in one of several holds of a cargo ship. In such cases, the size of a radioactive release resulting from a severe collision with another ship is determined not by the likelihood of compromising a single, robust package but by the probability that a certain fraction of 10's or 100's of individual packagings is compromised. The previous analysis (Spring, 1995) involved a statistical estimation of the frequency of accidents which would result in damage to a cask located in one of seven cargo holds in a collision with another ship. The results were obtained in the form of probabilities (frequencies) of accidents of increasing severity and of release fractions for each level of severity. This paper describes an extension of the same general method in which the multiple packages are assumed to be compacted by an intruding ship's bow until there is no free space in the hold. At such a point, the remaining energy of the colliding ship is assumed to be dissipated by progressively crushing the RAM packagings and the probability of a particular fraction of package failures is estimated by adaptation of the statistical method used previously. The parameters of a common, well-characterized packaging, the 6M with 2R inner containment vessel, were employed as an illustrative example of this analysis method. However, the method is readily applicable to other packagings for which crush strengths have been measured or can be estimated with satisfactory confidence. (authors)

Overview of core disruptive accidents

International Nuclear Information System (INIS)

Marchaterre, J.F.

1977-01-01

An overview of the analysis of core-disruptive accidents is given. These analyses are for the purpose of understanding and predicting fast reactor behavior in severe low probability accident conditions, to establish the consequences of such conditions and to provide a basis for evaluating consequence limiting design features. The methods are used to analyze core-disruptive accidents from initiating event to complete core disruption, the effects of the accident on reactor structures and the resulting radiological consequences are described

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)

PWR auxiliary systems, safety and emergency systems, accident analysis, operation

International Nuclear Information System (INIS)

Meyer, P.J.

1976-01-01

The author presents a description of PWR auxiliary systems like volume control, boric acid control, coolant purification, -degassing, -storage and -treatment system and waste processing systems. Residual heat removal systems, emergency systems and containment designs are discussed. As an accident analysis the author gives a survey over malfunctions and disturbances in the field of reactor operations. (TK) [de

Accidents - Chernobyl accident

International Nuclear Information System (INIS)

2004-01-01

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

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.

Underreporting of maritime accidents to vessel accident databases.

Science.gov (United States)

Hassel, Martin; Asbjørnslett, Bjørn Egil; Hole, Lars Petter

2011-11-01

Underreporting of maritime accidents is a problem not only for authorities trying to improve maritime safety through legislation, but also to risk management companies and other entities using maritime casualty statistics in risk and accident analysis. This study collected and compared casualty data from 01.01.2005 to 31.12.2009, from IHS Fairplay and the maritime authorities from a set of nations. The data was compared to find common records, and estimation of the true number of occurred accidents was performed using conditional probability given positive dependency between data sources, several variations of the capture-recapture method, calculation of best case scenario assuming perfect reporting, and scaling up a subset of casualty information from a marine insurance statistics database. The estimated upper limit reporting performance for the selected flag states ranged from 14% to 74%, while the corresponding estimated coverage of IHS Fairplay ranges from 4% to 62%. On average the study results document that the number of unreported accidents makes up roughly 50% of all occurred accidents. Even in a best case scenario, only a few flag states come close to perfect reporting (94%). The considerable scope of underreporting uncovered in the study, indicates that users of statistical vessel accident data should assume a certain degree of underreporting, and adjust their analyses accordingly. Whether to use correction factors, a safety margin, or rely on expert judgment, should be decided on a case by case basis. Copyright © 2011 Elsevier Ltd. All rights reserved.

Road accidents at night in the Netherlands : a national analysis according to official road accident data. Contribution to OECD Research Group TS 3 on Improving Road Safety at Night.

NARCIS (Netherlands)

Harris, S.

1979-01-01

The questionnaire about night-time accident data of the OECD Research Group TS 3 on Improving Road Safety at Night was filled in for the Netherlands. Thereafter a national analysis was written, using the already completed accident data questionnaire. Guidelines for the contents and presentation

Control rod drop accident analysis for the mixed core project in Ling Ao NPS

International Nuclear Information System (INIS)

Zhang Shishun; Zhou Zhou; Xiao Min

2004-01-01

AFA-2G assemblies in Ling Ao NPS (LNPS) have been replaced gradually by AFA-3G assemblies from cycle 2 and subsequent cycles. the enrichment of the fuels will be increased from 3.2% to 3.7% from cycle 3 in Ling Ao. Therefore, the study of ling Ao mixed core and increased enrichment have been performed since 2001. Lots of accidents need to be re-analyzed in Ling Ao NPS in order to verify its safety requirements for the new fuel management. Control rod drop accident for LNPS was re-analyzed in 2001 in frame of FRAMATOME ANP analytical methodology. The analytical codes used in the accident analysis include SCIENCE, ESPADON, CINEMA, CANTAL and FLICA III. The control rod drop accident analysis is performed with respect to the 10 reference cycles of the generic fuel management design for Ling Ao mixed core and increased enrichment study. The pre-drop FδH for the first transition cycles and other cycles are 1.52 and 1.55, respectively. For detected dropped rod configurations, the negative flux rate protection system actuates a reactor trip. For the non-detected dropped rod configurations, the minimum DNBR values have been evaluated with conservative analysis methodology and assumptions and the DNBR fuel design limit is respected the analytical results shows that, for all the non-detected dropped rod configurations, the minimum DNB margin is about 2% which occurs in AFA-2G fuel assembly in the first transition cycle. (author)

Response Analysis on Electrical Pulses under Severe Nuclear Accident Temperature Conditions Using an Abnormal Signal Simulation Analysis Module

Directory of Open Access Journals (Sweden)

Kil-Mo Koo

2012-01-01

Full Text Available Unlike design basis accidents, some inherent uncertainties of the reliability of instrumentations are expected while subjected to harsh environments (e.g., high temperature and pressure, high humidity, and high radioactivity occurring in severe nuclear accident conditions. Even under such conditions, an electrical signal should be within its expected range so that some mitigating actions can be taken based on the signal in the control room. For example, an industrial process control standard requires that the normal signal level for pressure, flow, and resistance temperature detector sensors be in the range of 4~20 mA for most instruments. Whereas, in the case that an abnormal signal is expected from an instrument, such a signal should be refined through a signal validation process so that the refined signal could be available in the control room. For some abnormal signals expected under severe accident conditions, to date, diagnostics and response analysis have been evaluated with an equivalent circuit model of real instruments, which is regarded as the best method. The main objective of this paper is to introduce a program designed to implement a diagnostic and response analysis for equivalent circuit modeling. The program links signal analysis tool code to abnormal signal simulation engine code not only as a one body order system, but also as a part of functions of a PC-based ASSA (abnormal signal simulation analysis module developed to obtain a varying range of the R-C circuit elements in high temperature conditions. As a result, a special function for abnormal pulse signal patterns can be obtained through the program, which in turn makes it possible to analyze the abnormal output pulse signals through a response characteristic of a 4~20 mA circuit model and a range of the elements changing with temperature under an accident condition.

Stepwise integral scaling method for severe accident analysis and its application to corium dispersion in direct containment heating

International Nuclear Information System (INIS)

Ishii, M.; Zhang, G.; No, H. C.; Eltwila, F.

1994-01-01

Accident sequences which lead to severe core damage and to possible radioactive fission products into the environment have a very low probability. However, the interest in this area increased significantly due to the occurrence of the small break loss-of-coolant accident at TMI-2 which led to partial core damage, and of the Chernobyl accident in the former USSR which led to extensive core disassembly and significant release of fission products over several countries. In particular, the latter accident raised the international concern over the potential consequences of severe accidents in nuclear reactor systems. One of the significant shortcomings in the analyses of severe accidents is the lack of well-established and reliable scaling criteria for various multiphase flow phenomena. However, the scaling criteria are essential to the severe accident, because the full scale tests are basically impossible to perform. They are required for (1) designing scaled down or simulation experiments, (2) evaluating data and extrapolating the data to prototypic conditions, and (3) developing correctly scaled physical models and correlations. In view of this, a new scaling method is developed for the analysis of severe accidents. Its approach is quite different from the conventional methods. In order to demonstrate its applicability, this new stepwise integral scaling method has been applied to the analysis of the corium dispersion problem in the direct containment heating. ((orig.))

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.

Reliability analysis of emergency decay heat removal system of nuclear ship under various accident conditions

International Nuclear Information System (INIS)

Matsuoka, Takeshi

1984-01-01

A reliability analysis is given for the emergency decay heat removal system of the Nuclear Ship ''Mutsu'' and the emergency sea water cooling system of the Nuclear Ship ''Savannah'', under ten typical nuclear ship accident conditions. Basic event probabilities under these accident conditions are estimated from literature survey. These systems of Mutsu and Savannah have almost the same reliability under the normal condition. The dispersive arrangement of a system is useful to prevent the reduction of the system reliability under the condition of an accident restricted in one room. As for the reliability of these two systems under various accident conditions, it is seen that the configuration and the environmental condition of a system are two main factors which determine the reliability of the system. Furthermore, it was found that, for the evaluation of the effectiveness of safety system of a nuclear ship, it is necessary to evaluate its reliability under various accident conditions. (author)

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.)

Analysis of Aircraft Crash Accident for WETF

International Nuclear Information System (INIS)

Jordan, Hans

2001-01-01

This report applies the methodology of DOE-STD-3014-96, ''Accident Analysis for Aircraft Crash into Hazardous Facilities'', to the Weapons Engineering Tritium Facility (WETF) at LANL. Straightforward application of that methodology shows that including local helicopter flights with those of all other aircraft with potential to impact the facility poses a facility impact risk slightly in excess of the DOE standard's threshold--10 -6 impacts per year. It is also shown that helicopters can penetrate the facility if their engines impact that facility's roof. However, a refinement of the helicopter impact analysis shows that penetration risk of the facility for all aircraft lies below the DOE standard's threshold. By that standard, therefore, the potential for release of hazardous material from the facility as a result of an aircraft crashing into the facility is negligible and need not be analyzed further

Reactor safety study. An assessment of accident risks in U.S. commercial nuclear power plants. Appendix I. Accident definition and use of event trees

International Nuclear Information System (INIS)

1975-10-01

Information is presented concerning accident definition and use of event trees, event tree methodology, potential accidents covered by the reactor safety study, analysis of potential accidents involving the reactor core, and analysis of potential accidents not involving the core

Analysis of heavy radiological accidents in NPP and gamma-irradiators

Energy Technology Data Exchange (ETDEWEB)

Angelov, V [Civil Defence Administration, Sofia (Bulgaria); Semova, T; Bonchev, Ts [Sofia Univ. (Bulgaria). Fizicheski Fakultet

1996-12-31

A review of several heavy radiological accidents, their cause, character, radioactivity emission, victims and economical impact is presented in the form of uniform tables. Eleven cases of incidents in power plants and 4 cases of accidents involving powerful gamma irradiators are considered. Radiological accidents in Bulgaria, not connected with the Kozloduy NPP, are listed. The human factor has been identified as the main cause for most of the accidents. It is stressed that the probability of heavy accident increases at the time of reactor refuelling, repair or testing. Technical failures could be eliminated by improved check and diagnostics procedures. 2 tabs., 12 refs.

Analysis of heavy radiological accidents in NPP and gamma-irradiators

International Nuclear Information System (INIS)

Angelov, V.; Semova, T.; Bonchev, Ts.

1995-01-01

A review of several heavy radiological accidents, their cause, character, radioactivity emission, victims and economical impact is presented in the form of uniform tables. Eleven cases of incidents in power plants and 4 cases of accidents involving powerful gamma irradiators are considered. Radiological accidents in Bulgaria, not connected with the Kozloduy NPP, are listed. The human factor has been identified as the main cause for most of the accidents. It is stressed that the probability of heavy accident increases at the time of reactor refuelling, repair or testing. Technical failures could be eliminated by improved check and diagnostics procedures. 2 tabs., 12 refs

Oxidation behavior analysis of cladding during severe accidents with combined codes for Qinshan Phase II Nuclear Power Plant

International Nuclear Information System (INIS)

Shi, Xingwei; Cao, Xinrong; Liu, Zhengzhi

2013-01-01

Highlights: • A new verified oxidation model of cladding has been added in Severe Accident Program (SAP). • A coupled analysis method utilizing RELAP5 and SAP codes has been developed and applied to analyze a SA caused by LBLOCA. • Analysis of cladding oxidation under a SA for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) has been performed by SAP. • Estimation of the production of hydrogen has been achieved by coupled codes. - Abstract: Core behavior at a high temperature is extremely complicated during transition from Design Basic Accident (DBA) to the severe accident (SA) in Light Water Reactors (LWRs). The progression of core damage is strongly affected by the behavior of fuel cladding (oxidation, embrittlement and burst). A Severe Accident Program (SAP) is developed to simulate the process of fuel cladding oxidation, rupture and relocation of core debris based on the oxidation models of cladding, candling of melted material and mechanical slumping of core components. Relying on the thermal–hydraulic boundary parameters calculated by RELAP5 code, analysis of a SA caused by the large break loss-of-coolant accident (LBLOCA) without mitigating measures for Qinshan Phase II Nuclear Power Plant (QSP-II NPP) was performed by SAP for finding the key sequences of accidents, estimating the amount of hydrogen generation and oxidation behavior of the cladding

Application of logical analysis of data to machinery-related accident prevention based on scarce data

International Nuclear Information System (INIS)

Jocelyn, Sabrina; Chinniah, Yuvin; Ouali, Mohamed-Salah; Yacout, Soumaya

2017-01-01

This paper deals with the application of Logical Analysis of Data (LAD) to machinery-related occupational accidents, using belt-conveyor-related accidents as an example. LAD is a pattern recognition and classification approach. It exploits the advancement in information technology and computational power in order to characterize the phenomenon under study. The application of LAD to machinery-related accident prevention is innovative. Ideally, accidents do not occur regularly, and as a result, companies have little data about them. The first objective of this paper is to demonstrate the feasibility of using LAD as an algorithm to characterize a small sample of machinery-related accidents with an adequate average classification accuracy. The second is to show that LAD can be used for prevention of machinery-related accidents. The results indicate that LAD is able to characterize different types of accidents with an average classification accuracy of 72–74%, which is satisfactory when compared with other studies dealing with large amounts of data where such a level of accuracy is considered adequate. The paper shows that the quantitative information provided by LAD about the patterns generated can be used as a logical way to prioritize risk factors. This prioritization helps safety practitioners make decisions regarding safety measures for machines. - Highlights: • LAD is presented as an innovative approach to prevent machinery-related accidents. • LAD is applied to a very small database of belt-conveyor-related accidents. • Despite scarce data, LAD generates patterns with adequate classification accuracy. • The patterns characterize different types of belt-conveyor-related accidents. • The patterns are useful to belt conveyor risk identification and risk estimation.

Analysis of the radiation accident in El Salvador

International Nuclear Information System (INIS)

Melara, N.E.

1998-01-01

On 5 February 1989 at 2 a.m. local time in a cobalt-60 industrial irradiation facility, a series of events started leading to one of the most serious radiation accidents in this type of installation. It took place in Soyapango, a city situated 5 km from San Salvador, the capital of the Republic of El Salvador. In this accident, three workers were involved in the first event and a further four in the second. When the accident took place, the activity level was approximately 0.66 PBq (18,000 Ci). The source became blocked when being lowered to its safe position, where upon the technician responsible for the irradiator entered the chamber in breach of the few inadequate safety procedures, accompanied by two colleagues from an adjacent department; the three workers suffered acute radiation exposure, with the result that one of them died six-and-a-half months later, the second had both his legs amputated at mid-thigh, while the third recovered completely. This article describes the irradiator, outlines the causes of the accident and analyses the economic and social repercussions, with the aim of helping teams responsible for radiation protection and safety in industrial irradiation facilities to identify potentially hazardous circumstances and avoid accidents. (author)

An analysis of severe air transport accidents

International Nuclear Information System (INIS)

McClure, J.D.; Luna, R.E.

1989-01-01

The objective of this paper is to analyze the severity of aircraft accidents that may involve the air transport of radioactive materials (RAM). One of the basic aims of this paper is to provide a numerical description of the severity of aircraft transport accidents so that the accident severity can be compared with the accident performance standards that are specified in IAEA Safety Series 6, the international packaging standards for the safe movement of RAM. The existing packaging regulations in most countries embrace the packaging standards developed by the IAEA. Historically, the packaging standards for Type B packages have been independent of the transport mode. That is, if the shipment occurs in a certified packaging, then the shipment can take place by any transport mode. In 1975, a legislative action occurred in the US Congress which led to the development of a package designed specifically for the air transport of plutonium. Changes were subsequently made to the US packaging regulations in 10CFR71 to incorporate the plutonium air transport performance standards. These standards were used to certify the air transport package for plutonium which is commonly referred to as PAT-1 (US NRC). The PAT-1 was certified by the US Nuclear Regulatory Commission in September 1978

10 CFR 76.85 - Assessment of accidents.

Science.gov (United States)

2010-01-01

... 10 Energy 2 2010-01-01 2010-01-01 false Assessment of accidents. 76.85 Section 76.85 Energy... Assessment of accidents. The Corporation shall perform an analysis of potential accidents and consequences to... postulated accidents which include internal and external events and natural phenomena in order to ensure...

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

International Nuclear Information System (INIS)

Borovkov, A.I.; Semenov, A.S.; Granovsky, V.S.; Kovtunova, S.V.

1995-01-01

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs

Finite element analysis of thermal stresses of the reactor vessel in a severe light water reactor accident

Energy Technology Data Exchange (ETDEWEB)

Borovkov, A.I.; Semenov, A.S. [St. Petersburg State Technical Univ. (Russian Federation); Granovsky, V.S.; Kovtunova, S.V. [Research Inst. of Technology, Sosnovy Bor (Russian Federation)

1995-12-31

The thermal stress and damage analysis of the light water reactor (LWR) vessel is considered in a severe accident conditions. The high temperature corium accumulates on the vessel bottom and necessary condition of its holding is intensive cooling of vessel. External flooding with outside cooling of the LWR vessel is one of the accident management strategies being proposed to ensure the integrity of the vessel after a severe accident. (author). 8 refs., 5 figs.

3-Dimensional Methodology for the Control Rod Ejection Accident Analysis Using UNICORN{sup TM}

Energy Technology Data Exchange (ETDEWEB)

Jang, Chan-su; Um, Kil-sup; Ahn, Dawk-hwan [Korea Nuclear Fuel Company, Taejon (Korea, Republic of); Kim, Yo-han; Sung, Chang-kyung [KEPRI, Taejon (Korea, Republic of); Song, Jae-seung [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2006-07-01

The control rod ejection accident has been analyzed with STRIKIN-II code using the point kinetics model coupled with conservative factors to address the three dimensional aspects. This may result in a severe transient with very high fuel enthalpy deposition. KNFC, under the support of KEPRI and KAERI, is developing 3-dimensional methodology for the rod ejection accident analysis using UNICORNTM (Unified Code of RETRAN, TORC and MASTER). For this purpose, 3-dimensional MASTER-TORC codes, which have been combined with the dynamic-link library by KAERI, are used in the transient analysis of the core and RETRAN code is used to estimate the enthalpy deposition in the hot rod.

Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study

Science.gov (United States)

AMIRI, Mehran; ARDESHIR, Abdollah; FAZEL ZARANDI, Mohammad Hossein

2014-01-01

Abstract Background The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. Methods The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were defined. The core of this work is devoted to analyzing the data from different perspectives such as age of workers, occupation and construction phase, day of the week, time of the day, seasonal analysis, regional considerations, type of accident, and body parts affected. Results Workers between 15-19 years old (TAR=13.4%) are almost six times more exposed to risk of accident than the average of all ages (TAR=2.51%). Laborers and structural workers (TAR=66.6%) and those working at heights (TAR=47.2%) experience more accidents than other groups of workers. Moreover, older workers over 65 years old (TSI=1.97%> average TSI=1.60%), work supervisors (TSI=12.20% >average TSI=9.09%), and night shift workers (TSI=1.89% >average TSI=1.47%) are more prone to severe accidents. Conclusion It is recommended that laborers, young workers, weekend and night shift workers be supervised more carefully in the workplace. Use of Personal Protective Equipment (PPE) should be compulsory in working environments, and special attention should be undertaken to people working outdoors and at heights. It is also suggested that policymakers pay more attention to the improvement of safety conditions in deprived and cold western regions. PMID:26005662

Risk-based Analysis of Construction Accidents in Iran During 2007-2011-Meta Analyze Study.

Science.gov (United States)

Amiri, Mehran; Ardeshir, Abdollah; Fazel Zarandi, Mohammad Hossein

2014-04-01

The present study aimed to investigate the characteristics of occupational accidents and frequency and severity of work related accidents in the construction industry among Iranian insured workers during the years 20072011. The Iranian Social Security Organization (ISSO) accident database containing 21,864 cases between the years 2007-2011 was applied in this study. In the next step, Total Accident Rate (TRA), Total Severity Index (TSI), and Risk Factor (RF) were defined. The core of this work is devoted to analyzing the data from different perspectives such as age of workers, occupation and construction phase, day of the week, time of the day, seasonal analysis, regional considerations, type of accident, and body parts affected. Workers between 15-19 years old (TAR=13.4%) are almost six times more exposed to risk of accident than the average of all ages (TAR=2.51%). Laborers and structural workers (TAR=66.6%) and those working at heights (TAR=47.2%) experience more accidents than other groups of workers. Moreover, older workers over 65 years old (TSI=1.97%> average TSI=1.60%), work supervisors (TSI=12.20% >average TSI=9.09%), and night shift workers (TSI=1.89% >average TSI=1.47%) are more prone to severe accidents. It is recommended that laborers, young workers, weekend and night shift workers be supervised more carefully in the workplace. Use of Personal Protective Equipment (PPE) should be compulsory in working environments, and special attention should be undertaken to people working outdoors and at heights. It is also suggested that policymakers pay more attention to the improvement of safety conditions in deprived and cold western regions.

Estimation of cost per severe accident for improvement of accident protection and consequence mitigation strategies

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2013-01-01

To assess the complex situations regarding the severe accidents such as what observed in Fukushima Accident, not only radiation protection aspects but also relevant aspects: health, environmental, economic and societal aspects; must be all included into the consequence assessment. In this study, the authors introduce the “cost per severe accident” as an index to analyze the consequences of severe accidents comprehensively. The cost per severe accident consists of various costs and consequences converted into monetary values. For the purpose of improvement of the accident protection and consequence mitigation strategies, the costs needed to introduce the protective actions, and health and psychological consequences are included in the present study. The evaluations of these costs and consequences were made based on the systematic consequence analysis using level 2 and 3 probabilistic safety assessment (PSA) codes. The accident sequences used in this analysis were taken from the results of level 2 seismic PSA of a virtual 1,100 MWe BWR-5. The doses to the public and the number of people affected were calculated using the level 3 PSA code OSCAAR of Japan Atomic Energy Agency (JAEA). The calculations have been made for 248 meteorological sequences, and the outputs are given as expectation values for various meteorological conditions. Using these outputs, the cost per severe accident is calculated based on the open documents on the Fukushima Accident regarding the cost of protective actions and compensations for psychological harms. Finally, optimized accident protection and consequence mitigation strategies are recommended taking into account the various aspects comprehensively using the cost per severe accident. The authors must emphasize that the aim is not to estimate the accident cost itself but to extend the scope of “risk-informed decision making” for continuous safety improvements of nuclear energy. (author)

Uncertainty analysis of accident notification time and emergency medical service response time in work zone traffic accidents.

Science.gov (United States)

Meng, Qiang; Weng, Jinxian

2013-01-01

Taking into account the uncertainty caused by exogenous factors, the accident notification time (ANT) and emergency medical service (EMS) response time were modeled as 2 random variables following the lognormal distribution. Their mean values and standard deviations were respectively formulated as the functions of environmental variables including crash time, road type, weekend, holiday, light condition, weather, and work zone type. Work zone traffic accident data from the Fatality Analysis Report System between 2002 and 2009 were utilized to determine the distributions of the ANT and the EMS arrival time in the United States. A mixed logistic regression model, taking into account the uncertainty associated with the ANT and the EMS response time, was developed to estimate the risk of death. The results showed that the uncertainty of the ANT was primarily influenced by crash time and road type, whereas the uncertainty of EMS response time is greatly affected by road type, weather, and light conditions. In addition, work zone accidents occurring during a holiday and in poor light conditions were found to be statistically associated with a longer mean ANT and longer EMS response time. The results also show that shortening the ANT was a more effective approach in reducing the risk of death than the EMS response time in work zones. To shorten the ANT and the EMS response time, work zone activities are suggested to be undertaken during non-holidays, during the daytime, and in good weather and light conditions.

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)

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

Background: 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. Methods: 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. Results: 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 106 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. Conclusion: 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. PMID:27057515

The application of the health effects models to the severe accident consequence analysis of nuclear power plants

International Nuclear Information System (INIS)

Yang Ling; Yeung, M.R.

1998-01-01

Health Effect Model (HEM) is an important model used in the analysis of severe accidents consequence of the Nuclear Power Plants (NPP). The accuracy of HEM affects the reliability of the assessment for the accidents consequences, and furthermore, the effectiveness of the emergency countermeasures taken for the health protection of the public around the NPPs. Based on the NUREG/CR4214 series reports, the paper sets appropriate parameters for HEM by studying both early and late HEMs used for domestic NPP accident consequence analysis. In the study, the Guangdong Daya Bay NPP is chosen as an example study to calculate the health risk of the Hong Kong population caused by Daya Bay NPP

A multivariate tobit analysis of highway accident-injury-severity rates.

Science.gov (United States)

Anastasopoulos, Panagiotis Ch; Shankar, Venky N; Haddock, John E; Mannering, Fred L

2012-03-01

Relatively recent research has illustrated the potential that tobit regression has in studying factors that affect vehicle accident rates (accidents per distance traveled) on specific roadway segments. Tobit regression has been used because accident rates on specific roadway segments are continuous data that are left-censored at zero (they are censored because accidents may not be observed on all roadway segments during the period over which data are collected). This censoring may arise from a number of sources, one of which being the possibility that less severe crashes may be under-reported and thus may be less likely to appear in crash databases. Traditional tobit-regression analyses have dealt with the overall accident rate (all crashes regardless of injury severity), so the issue of censoring by the severity of crashes has not been addressed. However, a tobit-regression approach that considers accident rates by injury-severity level, such as the rate of no-injury, possible injury and injury accidents per distance traveled (as opposed to all accidents regardless of injury-severity), can potentially provide new insights, and address the possibility that censoring may vary by crash-injury severity. Using five-year data from highways in Washington State, this paper estimates a multivariate tobit model of accident-injury-severity rates that addresses the possibility of differential censoring across injury-severity levels, while also accounting for the possible contemporaneous error correlation resulting from commonly shared unobserved characteristics across roadway segments. The empirical results show that the multivariate tobit model outperforms its univariate counterpart, is practically equivalent to the multivariate negative binomial model, and has the potential to provide a fuller understanding of the factors determining accident-injury-severity rates on specific roadway segments. Published by Elsevier Ltd.

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.

Offsite radiological consequence analysis for the bounding aircraft crash accident

International Nuclear Information System (INIS)

OBERG, B.D.

2003-01-01

The purpose of this calculation note is to quantitatively analyze a bounding aircraft crash accident for comparison to the DOE-STD-3009-94, ''Preparation Guide for U.S. Department of Energy Nonreactor Nuclear Facility Documented Safety Analyses'', Appendix A, Evaluation Guideline of 25 rem. The potential of aircraft impacting a facility was evaluated using the approach given in DOE-STD-3014-96, ''Accident Analysis for Aircraft Crash into Hazardous Facilities''. The following aircraft crash FR-equencies were determined for the Tank Farms in RPP-11736, ''Assessment Of Aircraft Crash FR-equency For The Hanford Site 200 Area Tank Farms'': (1) The total aircraft crash FR-equency is ''extremely unlikely.'' (2) The general aviation crash FR-equency is ''extremely unlikely.'' (3) The helicopter crash FR-equency is ''beyond extremely unlikely.'' (4) For the Hanford Site 200 Areas, other aircraft type, commercial or military, each above ground facility, and any other type of underground facility is ''beyond extremely unlikely.'' As the potential of aircraft crash into the 200 Area tank farms is more FR-equent than ''beyond extremely unlikely,'' consequence analysis of the aircraft crash is required

Integrated severe accident containment analysis with the CONTAIN computer code

International Nuclear Information System (INIS)

Bergeron, K.D.; Williams, D.C.; Rexroth, P.E.; Tills, J.L.

1985-12-01

Analysis of physical and radiological conditions iunside the containment building during a severe (core-melt) nuclear reactor accident requires quantitative evaluation of numerous highly disparate yet coupled phenomenologies. These include two-phase thermodynamics and thermal-hydraulics, aerosol physics, fission product phenomena, core-concrete interactions, the formation and combustion of flammable gases, and performance of engineered safety features. In the past, this complexity has meant that a complete containment analysis would require application of suites of separate computer codes each of which would treat only a narrower subset of these phenomena, e.g., a thermal-hydraulics code, an aerosol code, a core-concrete interaction code, etc. In this paper, we describe the development and some recent applications of the CONTAIN code, which offers an integrated treatment of the dominant containment phenomena and the interactions among them. We describe the results of a series of containment phenomenology studies, based upon realistic accident sequence analyses in actual plants. These calculations highlight various phenomenological effects that have potentially important implications for source term and/or containment loading issues, and which are difficult or impossible to treat using a less integrated code suite

Analysis of labor accidents in Brazil, 2004-2007 / Perfil dos acidentes de trabalho no Brasil, 2004/2007

OpenAIRE

Alves , Everton Fernando

2010-01-01

National audience; 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 laboraccidents were notified respectively in 2004 and 2007, with a trend towardsan increase in number. The state of Santa Catarina was the are...

Meta-analysis of the effect of road safety campaigns on accidents.

Science.gov (United States)

Phillips, Ross Owen; Ulleberg, Pål; Vaa, Truls

2011-05-01

A meta-analysis of 67 studies evaluating the effect of road safety campaigns on accidents is reported. A total of 119 results were extracted from the studies, which were reported in 12 different countries between 1975 and 2007. After allowing for publication bias and heterogeneity of effects, the weighted average effect of road safety campaigns is a 9% reduction in accidents (with 95% confidence that the weighted average is between -12 and -6%). To account for the variability of effects measured across studies, data were collected to characterise aspects of the campaign and evaluation design associated with each effect, and analysed to identify a model of seven campaign factors for testing by meta-regression. The model was tested using both fixed and random effect meta-regression, and dependency among effects was accounted for by aggregation. These analyses suggest positive associations between accident reduction and the use of personal communication or roadside media as part of a campaign delivery strategy. Campaigns with a drink-driving theme were also associated with greater accident reductions, while some of the analyses suggested that accompanying enforcement and short campaign duration (less than one month) are beneficial. Overall the results are consistent with the idea that campaigns can be more effective in the short term if the message is delivered with personal communication in a way that is proximal in space and time to the behaviour targeted by the campaign. Copyright © 2011 Elsevier Ltd. All rights reserved.

Cause finding experiments and environmental analysis on the accident of the fire and explosion in TRP bituminization facility

International Nuclear Information System (INIS)

Fujine, Sachio; Murata, Mikio; Abe, Hitoshi

1999-09-01

This report is the summary of the cause finding experiments and environmental analysis on the accident of the fire and explosion occurred at March 11th, 1997, in TRP bituminization facility of PNC (Power Reactor and Nuclear Fuel Development Corporation). Regarding the cause finding experiments, chemical components have been analyzed for the effluent samples taken from PNC's facility, bituminized mock waste has been produced using the simulated salt effluent prepared according to the results of chemical analysis, thermal analysis and experiment of runaway exothermic reaction have been conducted using the mock waste, and the component of flammable gases emitted from the heated waste have been collected and analyzed. Regarding environmental analysis on the accident, the amount of radioactive cesium released by the accident has been calculated by the comparative analysis using the atmospheric dispersion simulation code SPEEDI with the data of environmental monitoring and the public dose has been assessed. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Occupational accidents aboard merchant ships

DEFF Research Database (Denmark)

Hansen, H.L.; Nielsen, D.; Frydenberg, Morten

2002-01-01

Objectives: To investigate the frequency, circumstances, and causes of occupational accidents aboard merchant ships in international trade, and to identify risk factors for the occurrence of occupational accidents as well as dangerous working situations where possible preventive measures may...... be initiated. Methods: The study is a historical follow up on occupational accidents among crew aboard Danish merchant ships in the period 1993–7. Data were extracted from the Danish Maritime Authority and insurance data. Exact data on time at risk were available. Results: A total of 1993 accidents were...... aboard. Relative risks for notified accidents and accidents causing permanent disability of 5% or more were calculated in a multivariate analysis including ship type, occupation, age, time on board, change of ship since last employment period, and nationality. Foreigners had a considerably lower recorded...

Cernavoda CANDU severe accident evaluation

International Nuclear Information System (INIS)

Negut, G.; Marin, A.

1997-01-01

The papers present the activities dedicated to Romania Cernavoda Nuclear Power Plant first CANDU Unit severe accident evaluation. This activity is part of more general PSA assessment activities. CANDU specific safety features are calandria moderator and calandria vault water capabilities to remove the residual heat in the case of severe accidents, when the conventional heat sinks are no more available. Severe accidents evaluation, that is a deterministic thermal hydraulic analysis, assesses the accidents progression and gives the milestones when important events take place. This kind of assessment is important to evaluate to recovery time for the reactor operators that can lead to the accident mitigation. The Cernavoda CANDU unit is modeled for the of all heat sinks accident and results compared with the AECL CANDU 600 assessment. (orig.)

A proposal for accident management optimization based on the study of accident sequence analysis for a BWR

International Nuclear Information System (INIS)

Sobajima, M.

1998-01-01

The paper describes a proposal for accident management optimization based on the study of accident sequence and source term analyses for a BWR. In Japan, accident management measures are to be implemented in all LWRs by the year 2000 in accordance with the recommendation of the regulatory organization and based on the PSAs carried out by the utilities. Source terms were evaluated by the Japan Atomic Energy Research Institute (JAERI) with the THALES code for all BWR sequences in which loss of decay heat removal resulted in the largest release. Identification of the priority and importance of accident management measures was carried out for the sequences with larger risk contributions. Considerations for optimizing emergency operation guides are believed to be essential for risk reduction. (author)

Reference accident (Core disruption accident - safety analysis detailed report no. 11)

Energy Technology Data Exchange (ETDEWEB)

1988-01-15

The PEC safety analysis led to the conclusion that all credible sequences (incident sequences characterized by a frequency of occurrence above 10/sup minus 7/ events per year) are limited to the design basis conditions of components of the plant protection systems, and that none of them leads to a release of mechanical energy or to an extensive damage of the core and primary containment structures event in the case of failure to scram. Nevertheless, as is done in other countries for similar reactors, some events beyond the limits of credibility were considered for the PEC reactor. These were defined on a absolutely hypothetical basis that involves severe core disruption and dynamic loading of primary containment boundary. A series of containments, each having a different role, was designed to mitigate the radiological effects of a postulated core disruptive accident. The final aim was to demonstrate that residual heat can be removed and that the release of radioactivity to the environment is within acceptable limits.

Sensitivity analysis of thermal hydraulic response in containment at core meltdown accident

International Nuclear Information System (INIS)

Kobayashi, Kensuke; Ishigami, Tsutomu; Horii, Hideo; Chiba, Takemi.

1985-01-01

A sensitivity analysis of thermal hydraulic response in a containment during a 'station blackout' (the loss of all AC power) accident at Browns Ferry unit one plant was performed with the computer code MARCH 1.0. In the analysis, the plant station batteries were assumed to be available for 4h after the initiation of the accident. The thermal hydraulic response in the containment was calculated by varying several input data for MARCH 1.0 independently and the deviation among calculated results were investigated. The sensitivity analysis showed that (a) the containment would fail due to the overtemperature without any operator actions for plant recovery, which would be strongly dependent on the model of the debris-concrete interaction and the input parameters for specifying the containment failure modes in MARCH 1.0, (b) a core melting temperature and an amount of water left in a primary system at the end of the meltdown were identified as important parameters which influenced the time of the containment failure, and (c) experimental works regarding the parameters mentioned above could be recommended. (author)

Safety analysis results for cryostat ingress accidents in ITER

International Nuclear Information System (INIS)

Merrill, B.J.; Cadwallader, L.C.; Petti, D.A.

1996-01-01

Accidents involving the ingress of air or water into the cryostat of the International Thermonuclear Experimental Reactor (ITER) tokamak design have been analyzed with a modified version of the MELCOR code for the ITER Non-site Specific Safety Report (NSSR-1). The air ingress accident is the result of a postulated breach of the cryostat boundary into an adjoining room. MELCOR results for this accident demonstrate that the condensed air mass and increased heat loads are not a magnet safety concern, but that the partial vacuum in the adjoining room must be accommodated in the building design. The water ingress accident is the result of a postulated magnet arc that results in melting of a Primary Heat Transport System (PHTS) coolant pipe, discharging PHTS water and PHTS water activated corrosion products and HTO into the cryostat. MELCOR results for this accident demonstrate that the condensed water mass and increased heat loads are not a magnet safety concern, that the cryostat pressure remains below design limits, and that the corrosion product and HTO releases are well within the ITER release limits

Analysis of metal fuel transient overpower experiments with the SAS4A accident analysis code

International Nuclear Information System (INIS)

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

1990-01-01

The results of the SAS4A analysis of the M7 TREAT Metal fuel experiment are presented. New models incorporated in the metal fuel version of SAS4A are described. The computational results are compared with the experimental observations and this comparison is used in the interpretation of physical phenomena. This analysis was performed using the integrated metal fuel SAS4A version and covers a wide range of events, providing an increased degree of confidence in the SAS4A metal fuel accident analysis capabilities

Assessment of accident risks in the CRBRP. Volume 2. Appendices

Energy Technology Data Exchange (ETDEWEB)

None

1977-03-01

Appendices to Volume I include core-related accident-sequence definition, CRBRP risk-assessment sequence-probability determinations, failure-probability data, accident scenario evaluation, radioactive material release analysis, ex-core accident analysis, safety philosophy and design features, calculation of reactor accident consequences, sensitivity study, and risk from fires.

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

International Nuclear Information System (INIS)

Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

1995-01-01

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

Uncertainty and sensitivity analysis of food pathway results with the MACCS reactor accident consequence model

International Nuclear Information System (INIS)

Helton, J.C.; Johnson, J.D.; Rollstin, J.A.; Shiver, A.W.; Sprung, J.L.

1995-01-01

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

Detailed analysis of the TMI-2 accident scenario by using MARS/SCDAP

International Nuclear Information System (INIS)

Park, Rae Joon; Lee, Young Jin; Chung, Bub Dong

2009-01-01

As part of a benchmark analysis, the Three Mile Island Unit 2 (TMI-2) accident has been analyzed by using the MARS/SCDAP computer code. This analysis has been performed to estimate the efficiency of the MARS/SCDAP computer code and the predictive qualities of its models from an initiating event to a severe accident. The MARS/SCDAP results have shown that a reduction feed water to the steam generator caused the coolant to expand and initially increased the reactor coolant system (RCS) pressure. The pilot-operated relief valve (PORV) opened when the pressure reached 15.7 MPa, with a reactor scram occurring when the pressure reached 16.3 MPa. The PORV failed to close as the RCS pressure decreased, initiating a small break loss of coolant accident. The emergency core cooling was reduced by operators who thought that the pressurizer liquid level indicated a nearly full RCS, while coolant continued to be lost from the PORV. After an initial decrease in the RCS pressure, the pressurizer pressure remained at approximately 7 MPa. After a pump termination at 6,000 seconds, the liquid level in the reactor vessel decreased, which resulted in a core uncovery. Continued core degradation with a coolant boiling caused the pressurizer pressure to increase. The MARS/SCDAP results are very similar to the TMI-2 data

Coupled RELAP5/PANTHER/COBRA steam line break accident analysis in support of licensing DOEL 2 power uprate and steam generator replacement

International Nuclear Information System (INIS)

Zhang, J.; Bosso, S.; Henno, X.; Ouliddren, K.; Schneidesch, C.R.; Hove, W. van

2004-01-01

The nuclear reactor accident analyses using best estimate codes provide a better understanding and more accurate modeling of the key physical phenomena, which allows a more realistic evaluation of the conservatism and margins in the final safety analysis report (FSAR) accident analysis. The use of the best estimate codes and methods is necessary to meet the increasing technical, licensing and regulatory requirements for major plant changes (e.g. steam generator replacement), power uprate, core design optimization (cycle extension), as well as Periodic Safety Review. Since 1992, Tractebel Engineering (TE) has developed and applied a deterministic bounding approach to FASR accident analysis using the best estimate system thermal hydraulic code RELAP5/MOD2.5 and the subchannel thermal hydraulic code COBRA-3C. This approach has been accepted by the Belgian Safety Authorities, and turned out to be cost effective for most of the non-LOCA transient analyses. Since this approach adapts a decoupled modeling of the core responses, the analysis results often involved too large un-quantified conservatisms, due to either simplistic approximations for asymmetric accidents with strong 3D core neutronics - plant thermal hydraulics interactions, or additional penalties introduced from 'incoherent' initial/boundary conditions for separate plant and core analyses. Therefore, an external dynamic coupling between the RELAP5/MOD2.5 code and the 3-D neutronic code PANTHER was implemented since 1997 via the transient analysis code linkage program TALINK. Furthermore, a static linkage between the PANTHER code and the COBRA-3C code was developed for on-line calculation of (Departure from Nucleate Boiling Ratio (DNBR). TE intends to use the coupled code package for licensing non-symmetric FSAR accident analysis. The TE coupled code package has been applied to develop a main steam line break (MSLB) accident analysis methodology [using the TE deterministic bounding approach. The methodology

Bounding Accident Analysis for LLNL BSL-3 Facility

Energy Technology Data Exchange (ETDEWEB)

Johnson, Mark [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2010-09-14

The conclusion of this evaluation is that the consequence estimates in the EA can be reproduced using a public-accessible Gaussian plume-dispersion model and conservative modeling assumptions consistent with the accident scenario postulated in the EA. Also, the potential consequences to the public for the postulated accident would be far below the minimum infectious dose of one organism.

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

International Nuclear Information System (INIS)

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

1995-07-01

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

Cost per severe accident as an index for severe accident consequence assessment and its applications

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2014-01-01

The Fukushima Accident emphasizes the need to integrate the assessments of health effects, economic impacts, social impacts and environmental impacts, in order to perform a comprehensive consequence assessment of severe accidents in nuclear power plants. “Cost per severe accident” is introduced as an index for that purpose. The calculation methodology, including the consequence analysis using level 3 probabilistic risk assessment code OSCAAR and the calculation method of the cost per severe accident, is proposed. This methodology was applied to a virtual 1,100 MWe boiling water reactor. The breakdown of the cost per severe accident was provided. The radiation effect cost, the relocation cost and the decontamination cost were the three largest components. Sensitivity analyses were carried out, and parameters sensitive to cost per severe accident were specified. The cost per severe accident was compared with the amount of source terms, to demonstrate the performance of the cost per severe accident as an index to evaluate severe accident consequences. The ways to use the cost per severe accident for optimization of radiation protection countermeasures and for estimation of the effects of accident management strategies are discussed as its applications. - Highlights: • Cost per severe accident is used for severe accident consequence assessment. • Assessments of health, economic, social and environmental impacts are included. • Radiation effect, relocation and decontamination costs are important cost components. • Cost per severe accident can be used to optimize radiation protection measures. • Effects of accident management can be estimated using the cost per severe accident

A reactivity accidents simulation of the Fort Saint Vrain HTGR

International Nuclear Information System (INIS)

Fainer, Gerson

1980-01-01

A reactivity accidents analysis of the Fort Saint Vrain HTGR was made. The following accidents were analysed 1) A rod pair withdrawal accident during normal operation, 2) A rod pair ejection accident, 3) A rod pair withdrawal accident during startup operations at source levels and 4) Multiple rod pair withdrawal accident. All the simulations were performed by using the BLOOST-6 nuclear code The steady state reactor operation results obtained with the code were consistent with the design reactor data. The numerical analysis showed that all accidents - except the first one - cause particle failure. (author)

Analysis of accidents with organic material in health workers.

Science.gov (United States)

Vieira, Mariana; Padilha, Maria Itayra; Pinheiro, Regina Dal Castel

2011-01-01

This retrospective and descriptive study with a quantitative design aimed to evaluate occupational accidents with exposure to biological material, as well as the profile of workers, based on reporting forms sent to the Regional Reference Center of Occupational Health in Florianópolis/SC. Data collection was carried out through a survey of 118 reporting forms in 2007. Data were analyzed electronically. The occurrence of accidents was predominantly among nursing technicians, women and the mean age was 34.5 years. 73% of accidents involved percutaneous exposure, 78% had blood and fluid with blood, 44.91% resulted from invasive procedures. It was concluded that strategies to prevent the occurrence of accidents with biological material should include joint activities between workers and service management and should be directed at improving work conditions and organization.

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.

Prevention of radiation accidents and their consequences

International Nuclear Information System (INIS)

Khiski, J.

1976-01-01

Clearing out reasons for nuclear accidents enables to take effective measures to minimize them. The number of accidents in 1957 - 1974 is given. The frequency of accidents at various working places, while operating with various radioisotopes is presented. The analysis of accidents and the confirmation of these estimates can lead to the generalization of data and to the formulation of preventive measures [ru

Large LOCA accident analysis for AP1000 under earthquake

International Nuclear Information System (INIS)

Yu, Yu; Lv, Xuefeng; Niu, Fenglei

2015-01-01

Highlights: • Seismic failure event probability is induced by uncertainties in PGA and in Am. • Uncertainty in PGA is shared by all the components at the same place. • Relativity induced by sharing PGA value can be analyzed explicitly by MC method. • Multi components failures and accident sequences will occur under high PGA value. - Abstract: Seismic probabilistic safety assessment (PSA) is developed to give the insight of nuclear power plant risk under earthquake and the main contributors to the risk. However, component failure probability including the initial event frequency is the function of peak ground acceleration (PGA), and all the components especially the different kinds of components at same place will share the common ground shaking, which is one of the important factors to influence the result. In this paper, we propose an analysis method based on Monte Carlo (MC) simulation in which the effect of all components sharing the same PGA level can be expressed by explicit pattern. The Large LOCA accident in AP1000 is analyzed as an example, based on the seismic hazard curve used in this paper, the core damage frequency is almost equal to the initial event frequency, moreover the frequency of each accident sequence is close to and even equal to the initial event frequency, while the main contributors are seismic events since multi components and systems failures will happen simultaneously when a high value of PGA is sampled. The component failure probability is determined by uncertainties in PGA and in component seismic capacity, and the former is the crucial element to influence the result

Study On Safety Analysis Of PWR Reactor Core In Transient And Severe Accident Conditions

International Nuclear Information System (INIS)

Le Dai Dien; Hoang Minh Giang; Nguyen Thi Thanh Thuy; Nguyen Thi Tu Oanh; Le Thi Thu; Pham Tuan Nam; Tran Van Trung; Le Van Hong; Vo Thi Huong

2014-01-01

The cooperation research project on the Study on Safety Analysis of PWR Reactor Core in Transient and Severe Accident Conditions between Institute for Nuclear Science and Technology (INST), VINATOM and Korean Atomic Energy Research Institute (KAERI), Korea has been setup to strengthen the capability of researches in nuclear safety not only in mastering the methods and computer codes, but also in qualifying of young researchers in the field of nuclear safety analysis. Through the studies on the using of thermal hydraulics computer codes like RELAP5, COBRA, FLUENT and CFX the thermal hydraulics research group has made progress in the research including problems for safety analysis of APR1400 nuclear reactor, PIRT methodologies and sub-channel analysis. The study of severe accidents has been started by using MELCOR in collaboration with KAERI experts and the training on the fundamental phenomena occurred in postulated severe accident. For Vietnam side, VVER-1000 nuclear reactor is also intensively studied. The design of core catcher, reactor containment and severe accident management are the main tasks concerning VVER technology. The research results are presented in the 9 th National Conference on Mechanics, Ha Noi, December 8-9, 2012, the 10 th National Conference on Nuclear Science and Technology, Vung Tau, August 14-15, 2013, as well as published in the journal of Nuclear Science and Technology, Vietnam Nuclear Society and other journals. The skills and experience from using computer codes like RELAP5, MELCOR, ANSYS and COBRA in nuclear safety analysis are improved with the nuclear reactors APR1400, Westinghouse 4 loop PWR and especially the VVER-1000 chosen for the specific studies. During cooperation research project, man power and capability of Nuclear Safety center of INST have been strengthen. Three masters were graduated, 2 researchers are engaging in Ph.D course at Hanoi University of Science and Technology and University of Science and Technology, Korea

Epidemiology o.f· Traffic Accidents

African Journals Online (AJOL)

Accidents. An analysis of some 2 100 fatal traffic accidents gave the following results: males-79%; females-21%; a ratio of 4: 1. The high proportion of males to females killed in traffic accidents may be due to the fact that (a) more males commute daily in private and commercial vehicles;. (b) more females commute daily in ...

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

The potential risk of toxoplasmosis for traffic accidents: A systematic review and meta-analysis.

Science.gov (United States)

Gohardehi, Shaban; Sharif, Mehdi; Sarvi, Shahabeddin; Moosazadeh, Mahmood; Alizadeh-Navaei, Reza; Hosseini, Seyed Abdollah; Amouei, Afsaneh; Pagheh, Abdolsattar; Sadeghi, Mitra; Daryani, Ahmad

2018-06-12

Toxoplasmosis is a prevalent infectious disease. Although most people infected by Toxoplasma gondii are asymptomatic, evidence has suggested that this disease might affect some aspects of a host's behavior and associate with schizophrenia, suicide attempt, changes in various aspects of personality, and poor neurocognitive performance. These associations may play roles in increasing the risk of a number of incidents, such as traffic accidents, among infected people. In this regard, this study aimed to provide summary estimates for the available data on the potential risk of toxoplasmosis for traffic accidents. To this end, using a number of search terms, i.e. toxoplasmosis, Toxoplasma gondii, traffic accident, road accident, car accident, crash, and prevalence, literature searches (up to October 1, 2017) were carried out via 6 databases. The meta-analysis was conducted using the StatsDirect statistical software and a P-value less than 0.05 was regarded as significant in all statistical analyses. Out of 1841 identified studies, 9 studies were finally considered eligible for carrying out this systematic review. Reviewing results of these studies indicated that 5 out of 9 studies reported a significant relationship between Toxoplasma gondii and traffic accidents. Additionally, data related to gender showed significant differences between infected and control men and women. Considering age, reviewing the results of these studies revealed a significant difference between the infected people and the Toxoplasma-negative subjects under 45 years of age. However, no significant difference was found between the two groups aged 45 or older. Given these results, it can be concluded that Toxoplasma gondii significantly increases the risk of having traffic accidents. Copyright © 2018 Elsevier Inc. All rights reserved.

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

International Nuclear Information System (INIS)

1997-08-01

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

Accident management for severe accidents

International Nuclear Information System (INIS)

Bari, R.A.; Pratt, W.T.; Lehner, J.; Leonard, M.; Disalvo, R.; Sheron, B.

1988-01-01

The management of severe accidents in light water reactors is receiving much attention in several countries. The reduction of risk by measures and/or actions that would affect the behavior of a severe accident is discussed. The research program that is being conducted by the US Nuclear Regulatory Commission focuses on both in-vessel accident management and containment and release accident management. The key issues and approaches taken in this program are summarized. 6 refs

Event course analysis of core disruptive accidents

International Nuclear Information System (INIS)

Hering, W.; Homann, C.; Sengpiel, W.; Struwe, D.; Messainguiral, C.

1995-01-01

The theortical studies of the behavior of a PWR core in a meltdown accident are focused on hydrogen release, materials redistribution in the core area including forming of an oxide melt pool, quantity of melt and its composition, and temperatures attained by the RPV internals (esp. in the upper plenum) during the accident up to the time of melt relocation into the lower plenum. The calculations are done by the SCDAP/RELAP5 code. For its validation selected CORA results and Phebus FPTO results have been used. (orig.)

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

International Nuclear Information System (INIS)

Helton, J.C.; Johnson, J.D.; McKay, M.D.; Shiver, A.W.; Sprung, J.L.

1995-01-01

Uncertainty and sensitivity analysis techniques based on Latin hypercube sampling, partial correlation analysis and stepwise regression analysis are used in an investigation with the MACCS model of the early health effects associated with a severe accident at a nuclear power station. The primary purpose of this study is to provide guidance on the variables to be considered in future review work to reduce the uncertainty in the important variables used in the calculation of reactor accident consequences. The effects of 34 imprecisely known input variables on the following reactor accident consequences are studied: number of early fatalities, number of cases of prodromal vomiting, population dose within 10 mi of the reactor, population dose within 1000 mi of the reactor, individual early fatality probability within 1 mi of the reactor, and maximum early fatality distance. When the predicted variables are considered collectively, the following input variables were found to be the dominant contributors to uncertainty: scaling factor for horizontal dispersion, dry deposition velocity, inhalation protection factor for nonevacuees, groundshine shielding factor for nonevacuees, early fatality hazard function alpha value for bone marrow exposure, and scaling factor for vertical dispersion

Approach and results of the PWR low power and shutdown accident frequencies program - Coarse screening analysis for Surry

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Luckas, W.; Wong, S.M.; Fitzpatrick, R.G.

1991-01-01

Traditionally, probabilistic risk analyses of severe accidents in nuclear power plants have limited themselves to consideration of the set of initiating events occurring during full power operation. However, some analyses of accident initiators during low power, shutdown, and other modes of plant operation other than full power have been performed. These studies as well as the Chernobyl accident and recent operating experience at US pressurized water reactors (PWRs) suggested that risks during low power and shutdown could be significant. As such, the analysis of the frequencies, consequences, and risks of these accidents was identified as one task in the Nuclear Regulatory Commission staff's study of the implications of the Chernobyl accident to US commercial nuclear power plants. This program is an ongoing high priority effort at Brookhaven National Laboratory (BNL). The scope includes a Level 1 probabilistic risk assessment (PRA) with internal fire and flood for Surry Unit 1 (PWR). This program is also closely coupled to a parallel project for the Grand Gulf plant (BWWR) being conducted by SNL. The program is being performed in two phases. Phase 1 represents a coarse screening analysis to identify dominant accident scenarios as well as risk dominant plant configurations and plant operating states. In Phase 2, a detailed PRA will be performed for the dominant accident scenarios/operating states identified in Phase 1. The objectives, results and insights of Phase 1 are discussed in the paper

[Accidents in travellers - the hidden epidemic].

Science.gov (United States)

Walz, Alexander; Hatz, Christoph

2013-06-01

The risk of malaria and other communicable diseases is well addressed in pre-travel advice. Accidents are usually less discussed. Thus, we aimed at assessing accident figures for the Swiss population, based on data of the register from 2004 to 2008 of the largest Swiss accident insurance organization (SUVA). More than 139'000 accidents over 5 years showed that 65 % of the accidents overseas are injuries, and 24 % are caused by poisoning or harm by cold, heat or air pressure. Most accidents happened during leisure activities or sports. More than one third of the non-lethal and more than 50 % of the fatal accidents happened in Asia. More than three-quarters of non-lethal accidents take place in people between 25 and 54 years. One out of 74 insured persons has an accident abroad per year. Despite of many analysis short-comings of the data set with regard to overseas travel, the figures document the underestimated burden of disease caused by accidents abroad and should affect the given pre-health advice.

10 CFR 835.1304 - Nuclear accident dosimetry.