Core damage frequency estimation using accident sequence precursor data: 1990-1993

International Nuclear Information System (INIS)

Martz, H.F.

1998-01-01

The Nuclear Regulatory Commission's (NRC's) ongoing Accident Sequence Precursor (ASP) program uses probabilistic risk assessment (PRA) techniques to assess the potential for severe core damage (henceforth referred to simply as core damage) based on operating events. The types of operating events considered include accident sequence initiators, safety equipment failures, and degradation of plant conditions that could increase the probability that various postulated accident sequences occur. Such operating events potentially reduce the margin of safety available for prevention of core damage an thus can be considered as precursors to core damage. The current process for identifying, analyzing, and documenting ASP events is described in detail in Vanden Heuval et al. The significance of a Licensee Event Report (LER) event (or events) is measured by means of the conditional probability that the event leads to core damage, the so-called conditional core damage probability or, simply, CCDP. When the first ASP study results were published in 1982, it covered the period 1969--1979. In addition to identification and ranking of precursors, the original study attempted to estimate core damage frequency (CDF) based on the precursor events. The purpose of this paper is to compare the average annual CDF estimates calculated using the CCDP sum, Cooke-Goossens, Bier, and Abramson estimators for various reactor classes using the combined ASP data for the four years, 1990--1993. An important outcome of this comparison is an answer to the persistent question regarding the degree and effect of the positive bias of the CCDP sum method in practice. Note that this paper only compares the estimators with each other. Because the true average CDF is unknown, the estimation error is also unknown. Therefore, any observations or characterizations of bias are based on purely theoretical considerations

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal fires during mid-loop operations. Volume 3, Part 1, Main report

International Nuclear Information System (INIS)

Musicki, Z.; Chu, T.L.; Yang, J.; Ho, V.; Hou, Y.M.; Lin, J.; Siu, N.

1994-07-01

During l989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than fun power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in ' the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few. procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

ANS severe accident program overview & planning document

Energy Technology Data Exchange (ETDEWEB)

Taleyarkhan, R.P.

1995-09-01

The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10{sup {minus}6}/y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents.

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

MAAP - modular program for analyses of severe accidents

International Nuclear Information System (INIS)

Henry, R.E.; Lutz, R.J.

1990-01-01

The MAAP computer code was developed by Westinghouse as a fast, user-friendly, integrated analytical tool for evaluations of the sequences and consequences of severe accidents. The code allows a fully integrated treatment of thermohydraulic behavior and of the fission products in the primary system, the containment, and the ancillary buildings. This ensures interactive inclusion of all thermohydraulic events and of fission product behavior. All important phenomena which may occur in a major accident are contained in the modular code. In addition, many of the important parameters affecting the multitude of different phenomena can be defined by the user. In this way, it is possible to study the accuracy of the predicted course and of the consequences of a series of major accident phenomena. The MAAP code was subjected to extensive benchmarking with respect to the results of the experimental and theoretical programs, the findings obtained in other safety analyses using computers and data from accidents and transients in plants actually in operation. With the expected connection of the validation and test programs, the computer code attains a quality standard meeting the most stringent requirements in safety analyses. The code will be enlarged further in order to expand the number of benchmarks and the resolution of individual comparisons, and to ensure that future MAAP models will be in better agreement with the experiments and experiences of industry. (orig.) [de

ANS severe accident program overview ampersand planning document

International Nuclear Information System (INIS)

Taleyarkhan, R.P.

1995-09-01

The Advanced Neutron Source (ANS) severe accident document was developed to provide a concise and coherent mechanism for presenting the ANS SAP goals, a strategy satisfying these goals, a succinct summary of the work done to date, and what needs to be done in the future to ensure timely licensability. Guidance was received from various bodies [viz., panel members of the ANS severe accident workshop and safety review committee, Department of Energy (DOE) orders, Nuclear Regulatory Commission (NRC) requirements for ALWRs and advanced reactors, ACRS comments, world-wide trends] were utilized to set up the ANS-relevant SAS goals and strategy. An in-containment worker protection goal was also set up to account for the routine experimenters and other workers within containment. The strategy for achieving the goals is centered upon closing the severe accident issues that have the potential for becoming certification issues when assessed against realistic bounding events. Realistic bounding events are defined as events with an occurrency frequency greater than 10 -6 /y. Currently, based upon the level-1 probabilistic risk assessment studies, the realistic bounding events for application for issue closure are flow blockage of fuel element coolant channels, and rapid depressurization-related accidents

The radiological consequences of degraded core accidents for the Sizewell PWR The impact of adopting revised frequencies of occurrence

CERN Document Server

Kelly, G N

1983-01-01

The radiological consequences of degraded core accidents postulated for the Sizewell PWR were assessed in an earlier study and the results published in NRPB-R137. Further analyses have since been made by the Central Electricity Generating Board (CEGB) of degraded core accidents which have led to a revision of their predicted frequencies of occurrence. The implications of these revised frequencies, in terms of the risk to the public from degraded core accidents, are evaluated in this report. Increases, by factors typically within the range of about 1.5 to 7, are predicted in the consequences, compared with those estimated in the earlier study. However, the predicted risk from degraded core accidents, despite these increases, remains exceedingly small.

Review of the severe accident risk reduction program (SARRP) containment event trees

International Nuclear Information System (INIS)

1986-05-01

A part of the Severe Accident Risk Reduction Program, researchers at Sandia National Laboratories have constructed a group of containment event trees to be used in the analysis of key accident sequences for light water reactors (LWR) during postulated severe accidents. The ultimate goal of the program is to provide to the NRC staff a current assessment of the risk from severe reactor accidents for a group of five light water reactors. This review specifically focuses on the development and construction of the containment event trees and the results for containment failure probability, modes and timing. The report first gives the background on the program, the review criteria, and a summary of the observations, findings and recommendations. secondly, the individual reviews of each committee member on the event trees is presented. Finally, a review is provided on the computer model used to construct and evaluate the event trees

Program for accident and incident management support, AIMS

International Nuclear Information System (INIS)

Putra, M.A.

1993-12-01

A prototype of an advisory computer program is presented which could be used in monitoring and analyzing an ongoing incident in a nuclear power plant. The advisory computer program, called the Accident and Incident Management Support (AIMS), focuses on processing a set of data that is to be transmitted from a nuclear power plant to a national or regional emergency center during an incident. The AIMS program will assess the reactor conditions by processing the measured plant parameters. The applied model of the power plant contains a level of complexity that is comparable with the simplified plant model that the power plant operator uses. A standardized decay heat function and a steam water property library is used in the integral balance equations for mass and energy. A simulation of the station blackout accident of the Borssele plant is used to test the program. The program predicts successively: (1) the time of dryout of the steam generators, (2) the time of saturation of the primary system, and (3) the onset of core uncovery. The coolant system with the actual water levels will be displayed on the screen. (orig./HP)

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.

1994-06-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices A--D. Volume 2, Part 2

Energy Technology Data Exchange (ETDEWEB)

Chu, T.L.; Musicki, Z.; Kohut, P. [Brookhaven National Lab., Upton, NY (United States)] [and others

1994-06-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the Potential risks during low Power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the Plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful. This document, Volume 2, Pt. 2 provides appendices A through D of this report.

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

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 7--12). Volume 2, Part 1B

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.

1994-06-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specific shutdown accidents would be useful

REAC/TS radiation accident registry. Update of accidents in the United States

International Nuclear Information System (INIS)

Ricks, R.C.; Berger, M.E.; Holloway, E.C.; Goans, R.E.

2000-01-01

Serious injury due to ionizing radiation is a rare occurrence. From 1944 to the present, 243 US accidents meeting dose criteria for classification as serious are documented in the REAC/TS Registry. Thirty individuals have lost their lives in radiation accidents in the United States. The Registry is part of the overall REAC/TS program providing 24-hour direct or consultative assistance regarding medical and heath physics problems associated with radiation accidents in local, national, and international incidents. The REAC/TS Registry serves as a repository of medically important information documenting the consequences of these accidents. Registry data are gathered from various sources. These include reports from the World Heath Organization (WHO), International Atomic Energy Agency (IAEA), US Nuclear Regulatory Commission (US NRC), state radiological health departments, medical/health physics literature, personal communication, the Internet, and most frequently, from calls for medical assistance to REAC/TS, as part of our 24-hour medical assistance program. The REAC/TS Registry for documentation of radiation accidents serves several useful purposes: 1) weaknesses in design, safety practices, training or control can be identified, and trends noted; 2) information regarding the medical consequences of injuries and the efficacy of treatment protocols is available to the treating physician; and 3) Registry case studies serve as valuable teaching tools. This presentation will review and summarize data on the US radiation accidents including their classification by device, accident circumstances, and frequency by respective states. Data regarding accidents with fatal outcomes will be reviewed. The inclusion of Registry data in the IAEA's International Reporting System of Radiation Events (RADEV) will also be discussed. (author)

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

International Nuclear Information System (INIS)

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

2010-01-01

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

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.

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Appendices E (Sections E.1--E.8). Volume 2, Part 3A

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.

1994-06-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. The authors recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown. Procedures written specifically for shutdown accidents would be useful

Analysis of rail accident frequencies and severities for the assessment of radioactive material transport risk - Summary report

International Nuclear Information System (INIS)

Heywood, J.D.; Schwartz, G.; Fett, J.

2001-01-01

This shortened version of the final contractual report to the European Commission DGXVII summarises the work performed and the conclusions drawn from consideration, comparison and analysis of transport accident frequency and severity assessment methods for radioactive material transport by rail. This paper aims to provide an introduction to the study whose final report is 155 pages in length. The findings are based on a comprehensive review of transport risk assessment methods and related databases available to EU member states. The emphasis has been on the probabilistic accident severity and frequency assessment methodologies developed and used by the organisations involved in this EU-funded research project - AEA Technology and GRS. The results should be of major assistance in the understanding and development of standardised quantitative risk assessment models. Further work is suggested to underpin the development of a harmonised accident methodology including the collection of more detailed rail data and analysis on a year by year basis as well as further consideration of the assumptions made for fire accident scenarios. (author)

Aviation Safety Program: Weather Accident Prevention (WxAP) Project Overview and Status

Science.gov (United States)

Nadell, Shari-Beth

2003-01-01

This paper presents a project overview and status for the Weather Accident Prevention (WxAP) aviation safety program. The topics include: 1) Weather Accident Prevention Project Background/History; 2) Project Modifications; 3) Project Accomplishments; and 4) Project's Next Steps.

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations, Main report (Chapters 1--6). Volume 2, Part 1A

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.

1992-06-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this report is to document the approach utilized in the Surry plant and discuss the results obtained. A parallel report for the Grand Gulf plant is prepared by SNL. This study shows that the core-damage frequency during mid-loop operation at the Surry plant is comparable to that of power operation. We recognize that there is very large uncertainty in the human error probabilities in this study. This study identified that only a few procedures are available for mitigating accidents that may occur during shutdown written specifically for shutdown accidents would be useful. This document presents Chapters 1--6 of the report

Driver behavior and accident frequency in school zones: Assessing the impact of sign saturation.

Science.gov (United States)

Strawderman, Lesley; Rahman, Md Mahmudur; Huang, Yunchen; Nandi, Apurba

2015-09-01

Based on the models of human information processing, if a driver observes too many of the same signs, he or she may no longer pay attention to those signs. In the case of school zones, this expected effect may lead to non-compliance to posted speeds, negatively impacting safety around nearby schools. This study aims to investigate the effect of the number of nearby school zones on driver behavior (vehicle speed and compliance) and accident frequency. As a measure of the density of school zones, this study introduced and defined a new term sign saturation and presented a methodology to calculate sign saturation for school zones. Results found a significant effect of sign saturation on vehicle speed, compliance, and accident frequency. This study also examined the speeding behavior in school zones for different time of the day and day of the week. Results found that speeding was more prevalent in the early mornings and during the weekends. Copyright © 2015 Elsevier Ltd. All rights reserved.

Methodological aspects of core meltdown accidents frequency estimates

International Nuclear Information System (INIS)

Matthis, P.

1984-01-01

A survey is given of the work of the ecological institute relating to models and methods used in the German Risk Study for the assessment of core meltdown accident frequency. A statistical model used by the ecological institute for the estimation of the outage behaviour of components is taken as a comparison, which leads to the conclusion that no appropriate methods for the assessment of component reliability are available to date. Furthermore, there are no secured methods for error propagation computation. The lower limits for the ranges of reliability of components are calculated by approximation. As a result of imperfect modelling and of a number of methodical inaccuracies and neglects, the German Risk Study underestimates the ranges of component reliability by a factor of 3 to 70 (depending on the type of component). (RF) [de

Drivers of accident preparedness and safety: evidence from the RMP Rule

International Nuclear Information System (INIS)

Kleindorfer, Paul R.; Elliott, Michael R.; Wang Yanlin; Lowe, Robert A.

2004-01-01

This paper provides an overview of recent results derived from the accident history data collected under 112(r) of the Clean Air Act Amendments (the Risk Management Program (RMP) Rule) covering the period 1994-2000, together with a preliminary assessment of the effectiveness of the RMP Rule as a form of Management System Regulation. These were undertaken at the University of Pennsylvania by a multi-disciplinary team of economists, statisticians and epidemiologists with the support of the US Environmental Protection Agency and its Office of Emergency Prevention, Preparedness and Response (OEPPR, formerly CEPPO). Section 112(r) of the Clean Air Act Amendments of 1990 requires that chemical facilities in the US that had on premises more than specified quantities of toxic or flammable chemicals file a 5-year history of accidents. The initial data reported under the RMP Rule covered roughly the period from mid-1994 through mid-2000, and provided details on economic, environmental and acute health affects resulting from accidents at some 15,000 US chemical facilities for this period. This paper reviews research based on this data. The research is in the form of a retrospective cohort study that considers the statistical associations between accident frequency and accident severity at covered facilities (the outcome variables of interest) and a number of facility characteristics (the available predictor variables provided by the RMP Rule), the latter including such facility characteristics as size, hazardousness, financial characteristics of parent company-owners of the facility, regulatory programs in force at the facility, and host community characteristics for the surrounding county in which the facility was located, as captured in the 1990 Census. Among the findings reviewed are: (1) positive associations with (a measure of) facility hazardousness and accident, injury and economic costs of accidents; (2) positive (resp., negative) associations between accident

[Multicenter paragliding accident study 1990].

Science.gov (United States)

Lautenschlager, S; Karli, U; Matter, P

1992-01-01

During the period from 1.1.90 until 31.12.90, 86 injuries associated with paragliding were analyzed in a prospective study in 12 different Swiss hospitals with reference to causes, patterns, and frequencies. The injuries showed a mean score of over 2 and were classified as severe. Most frequent spine injuries (36%) and lesions of the lower extremity (35%) with a high risk of the ankles were diagnosed. One accident was fatal. 60% of the accidents happened during landing, 26% during launching and 14% during flight. Half of the pilots were affected during their primary training course. Most accidents were caused by inflight error of judgement--especially incorrect estimation of wind conditions--and further the choice of unfavourable landing sites. In contrast to previous injury-reports, only one equipment failure could be noted, but often the equipment was not corresponding with the experience and the weight of the pilot. To reduce the frequency of paragliding-injuries an accurate choice of equipment and an increased attention to environmental factors is mandatory. Furthermore an education-program regarding the attitude and intelligence of the pilot should be included in training courses.

Investigation of Qom Rural Area Water Network Accident in 2010 and Minimization Approaches of Accident Frequencies

Directory of Open Access Journals (Sweden)

Hossein Jafari Mansoorian

2016-02-01

Full Text Available Background & Aims of the Study : Accidents in water networks can lead to increase the uncounted water, costs of repair, maintenance, restoration and enter water contaminants to water network. The aim of this study is to survey the accidents of Qom rural water network and choose the right approaches to reduce the number of accidents. Materials & Methods: In this cross-sectional study, four sector of Qom province (Markazi, Dastjerd, Kahak and Qahan, were assessed over a period of 8 months (July – January 2010. This study was conducted through questionnaire of Ministry of Energy. Results: The total number of accidents was 763. The highest number of accidents in the four sectors was related to Markazi sector with 228 accidents. According to the time of the accident, the highest and lowest number of accident was related to September (19.7% and November (6.8%, respectively. According to the location of the accident on network, the highest and lowest number of accident was related to distribution network (64% and connections (17.5% and transmission pipe (18.34%, respectively. According to the type of the accident, the highest and lowest number of accident was related to breaking (47.8% and gasket failure (1.2%, respectively. Considering with the pipes’ material, the highest and lowest number of accident was related to polyethylene pipes (93% and steel and cast iron pipes (0.5%, 0.5%, respectively. Conclusions: Due to the high break rate of Polyethylene pipes, it is recommended to be placed in priority of leak detection and rehabilitation.   .

The program of international intercomparison of accident dosimetry; Le programme d'intercomparaison internationale de dosimetrie d'accident 10-12 juin 2002

Energy Technology Data Exchange (ETDEWEB)

NONE

2002-06-01

The French institute of radioprotection and nuclear safety (IRSN) has carried out in June 2002 an international intercomparison program for the testing of the physical and biological accident dosimetry techniques. The intercomparison is jointly organized by the IRSN and the OECD-NEA with the sustain of the European commission and the collaboration of the CEA centre of Valduc (France). About 30 countries have participated to this program. Each country has supplied its own dosimeters and biological samples which have been irradiated using the Silene reactor of CEA-Valduc or a {sup 60}Co source. These experiments allow to test the new dosimetric techniques that have been developed since the previous intercomparison program (1993) and to confirm or improve the performances of older techniques. Aside from the intercomparison exercise, this report makes a status of the known radiological accidents and of the effects of high doses of ionizing radiations on human health (symptoms, therapeutics). It explains the phenomenology of criticality accidents, the prevention means, and the history of such accidents up to the Tokai-Mura one in 1999. Finally, the dosimetry of criticality is presented with its physical and biological techniques. (J.S.)

The Nordic safety program on accident consequence assessment

International Nuclear Information System (INIS)

Tveten, U.

1988-01-01

One important part of Nordic cooperation is partially funded by the Nordic Council of Ministers, namely the work performed within the Nordic Safety Program (often referred to as the NKA projects). NKA is the Nordic abbreviation of the Nordic Liaison Committee on Atomic Energy. One program area in the present four-year period is concerned with problems related to reactor accident consequence assessment, and contains almost twenty projects covering a wide range of subjects. The author is program coordinator for this program area. The program will be completed in 1989. The program was strongly influenced by Chernobyl, and a number of new projects were included in the program in 1986. Involved in the program are these Nordic institutions: Riso National Laboratory (Denmark). Technical Research Centre of Finland. Finnish Centre for Radiation and Nuclear Safety. Finnish Meteorological Institute. Institute for Energy Technology (Norway). Agricultural University of Norway. Meteorological Institute of Norway. Studsvik Energiteknik AB (Sweden). National Defence Research Laboratory (Sweden)

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

Energy Technology Data Exchange (ETDEWEB)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States); Holmes, B. [AEA Technology, Dorset (United Kingdom)] [and others

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis.

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix I, Volume 2, Part 5

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Lab. (BNL) and Sandia National Labs. (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The objective of this volume of the report is to document the approach utilized in the level-1 internal events PRA for the Surry plant, and discuss the results obtained. A phased approach was used in the level-1 program. In phase 1, which was completed in Fall 1991, a coarse screening analysis examining accidents initiated by internal events (including internal fire and flood) was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis

User Interface for the SMAC Traffic Accident Reconstruction Program

Directory of Open Access Journals (Sweden)

Rok Krulec

2003-11-01

Full Text Available This paper describes the development of the user interfacefor the traffic accident reconstruction program SMAC. Threebasic modules of software will be presented. Initial parametersinput and visualization, using graphics library for simulation of3D space, which form a graphical user interface, will be explainedin more detail. The modules have been developed usingdifferent technologies and programming approaches to increaseflexibility in further development and to take maximumadvantage of the currently accessible computer hardware, sothat module to module communication is also mentioned.

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

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

Final results of the Cuban program for children from areas affected by the Chernobyl accident

International Nuclear Information System (INIS)

Medina, Julio; García, Omar

2016-01-01

From 1990 till 2011, a free health care program for children from areas affected by the Chernobyl accident was organized and implemented by the Cuban Government. A significant number of qualified personal in medicine and science collaborated with this program. The information gathered for more than 21 years and the conclusive results obtained in the program are presented here. During this time, 26 114 patients, 84 % children, mostly from Ukraine, Russia and Belarus were treated in the mentioned program. Other aspects of the Cuban medical assistance in Ukraine and the main results obtained in the radiological impact evaluation of the accident on the children treated in Cuba are also included. (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.

Quantitative risk trends deriving from PSA-based event analyses. Analysis of results from U.S.NRC's accident sequence precursor program

International Nuclear Information System (INIS)

Watanabe, Norio

2004-01-01

The United States Nuclear Regulatory Commission (U.S.NRC) has been carrying out the Accident Sequence Precursor (ASP) Program to identify and categorize precursors to potential severe core damage accident sequences using the probabilistic safety assessment (PSA) technique. The ASP Program has identified a lot of risk significant events as precursors that occurred at U.S. nuclear power plants. Although the results from the ASP Program include valuable information that could be useful for obtaining and characterizing risk significant insights and for monitoring risk trends in nuclear power industry, there are only a few attempts to determine and develop the trends using the ASP results. The present study examines and discusses quantitative risk trends for the industry level, using two indicators, that is, the occurrence frequency of precursors and the annual core damage probability, deriving from the results of the ASP analysis. It is shown that the core damage risk at U.S. nuclear power plants has been lowered and the likelihood of risk significant events has been remarkably decreasing. As well, the present study demonstrates that two risk indicators used here can provide quantitative information useful for examining and monitoring the risk trends and/or risk characteristics in nuclear power industry. (author)

Cerebrovascular accidents complicating transcatheter aortic valve implantation: frequency, timing and impact on outcomes.

Science.gov (United States)

Stortecky, Stefan; Windecker, Stephan; Pilgrim, Thomas; Heg, Dik; Buellesfeld, Lutz; Khattab, Ahmed A; Huber, Christoph; Gloekler, Steffen; Nietlispach, Fabian; Mattle, Heinrich; Jüni, Peter; Wenaweser, Peter

2012-05-15

Cerebrovascular accidents (CVA) are considered among the most serious adverse events after transcatheter aortic valve implantation (TAVI). The objective of the present study was to evaluate the frequency and timing of CVA after TAVI and to investigate the impact on clinical outcomes within 30 days of the procedure. Between August 2007 and October 2011, 389 high-risk elderly patients with symptomatic severe aortic stenosis underwent TAVI via transfemoral, transapical or subclavian access. A total of 14 patients (3.6%) experienced at least one CVA within 30 days of follow-up and most events (74%) occurred within the first day of the procedure. Patients with CVA had an increased risk of all-cause (42.3% vs. 5.1%, ORadjusted 11.7, 95% CI 3.4-40.3, pCerebrovascular accidents among patients undergoing TAVI occur predominantly during the periprocedural period, are associated with multiple implantation attempts of the bioprosthesis and significantly impair prognosis.

Cooperative Severe Accident Research Program of the USNRC and its foreign partners: Program content and principal results

International Nuclear Information System (INIS)

Wright, R.W.; Eltawila, F.

1993-01-01

The U.S. Nuclear Regulatory Commission (NRC) and its associated foreign partners have been engaged in an extensive Cooperative Severe Accident Research Program. In addition to the NRC, the partners currently include Belgium, the Czech Republic, Canada, Finland, France, Germany, Hungary, Italy, Japan, Korea, the Netherlands, Russia, Spain, Sweden, Switzerland, Taiwan, the United Kingdom, and the Community of European Countries. The purpose of this research is to provide a technical basis for decisions involved in potential severe accidents in light water reactor (LWR) power plants. The research includes relatively large-scale integral tests and smaller scale separate-effects experiments on the dominant phenomena regarding severe accident behavior in LWR power plants, the development of phenomenological models of the key phenomena involved, and the development and validation of large computer codes for use in the analysis of core behavior and of a LWR systems behavior under severe accident conditions. The research results are also used in probabilistic risk assessment for LWRS

Historical aspects of radiation accidents

International Nuclear Information System (INIS)

Mettler, F.A. Jr.; Ricks, R.C.

1990-01-01

Radiation accidents are extremely rare events; however, the last two years have witnessed the largest radiation accidents in both the eastern and western hemispheres. It is the purpose of this chapter to review how radiation accidents are categorized, examine the temporal changes in frequency and severity, give illustrative examples of several types of radiation accidents, and finally, to describe the various registries for radiation accidents

France-Japan collaboration on the severe accident studies for ASTRID. Outcomes and future work program

International Nuclear Information System (INIS)

Serre, F.; Bertrand, F.; Bachrata, A.; Marie, N.; Kubo, Shigenobu; Kamiyama, Kenji; Carluec, B.; Farges, B.; Koyama, K.

2017-01-01

The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator of GenIV sodium-cooled fast reactor (SFR) designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are first to prevent the core melting, in particular by the development of an innovative core (named CFV core) with low void worth and complementary safety prevention devices, and second, to enhance the reactor resistance to severe accidents by design. In order to mitigate the consequences of hypothetical core melting situations, specific provisions (mitigation devices) are added to the core and to the reactor. To meet these ASTRID objectives, a large R and D program was launched in the Severe Accident domain by the CEA, with collaboration of AREVA NP, JAEA, MFBR and MHI organizations, in the frame of the France-Japan ASTRID and SFRs collaboration agreement. This R and D program covers exchanges on severe accident conditions to be studied for the SFR safety cases, the methodology to study these situations, ASTRID severe accident simulations, the comparison and understanding of the ASTRID and JSFR reactor behavior under these situations, the development and adaptation of simulation tools, and, despite an already large existing experimental database, a complementary experimental program to improve the knowledge and reduce the uncertainties. This paper will present the collaboration work performed on the Severe Accidents studies. (author)

Addressing severe accidents in the CANDU 9 design

International Nuclear Information System (INIS)

Nijhawan, S.M.; Wight, A.L.; Snell, V.G.

1998-01-01

CANDU 9 is a single-unit evolutionary heavy-water reactor based on the Bruce/Darlington plants. Severe accident issues are being systematically addressed in CANDU 9, which includes a number of unique features for prevention and mitigation of severe accidents. A comprehensive severe accident program has been formulated with feedback from potential clients and the Canadian regulatory agency. Preliminary Probabilistic Safety Analyses have identified the sequences and frequency of system and human failures that may potentially lead to initial conditions indicating onset of severe core damage. Severe accident consequence analyses have used these sequences as a guide to assess passive heat sinks for the core, and containment performance. Estimates of the containment response to mass and energy injections typical of postulated severe accidents have been made and the results are presented. We find that inherent CANDU severe accident mitigation features, such as the presence of large water volumes near the fuel (moderator and shield tank), permit a relatively slow severe accident progression under most plant damage states, facilitate debris coolability and allow ample time for the operator to arrest the progression within, progressively, the fuel channels, calandria vessel or shield tank. The large-volume CANDU 9 containment design complements these features because of the long times to reach failure

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1. Volume 5: Analysis of core damage frequency from seismic events during mid-loop operations

International Nuclear Information System (INIS)

Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

1994-08-01

In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1) and the other at Sandia National Laboratories studying a boiling water reactor (Grand Gulf). Both the Brookhaven and Sandia projects have examined only accidents initiated by internal plant faults--so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling shutdown conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Surry Unit 1. All of the many systems modeling assumptions, component non-seismic failure rates, and human error rates that were used in the internal-initiator study at Surry have been adopted here, so that the results of the two studies can be as comparable as possible. Both the Brookhaven study and this study examine only two shutdown plant operating states (POSs) during refueling outages at Surry, called POS 6 and POS 10, which represent mid-loop operation before and after refueling, respectively. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POSs 6 and 10. The results of the analysis are that the core-damage frequency of earthquake-initiated accidents during refueling outages in POS 6 and POS 10 is found to be low in absolute terms, less than 10 -6 /year

Annual technical meeting of the NRC cooperative severe accident research program

International Nuclear Information System (INIS)

Silver, E.G.

1993-01-01

This brief report summarizes the 1992 annual technical meeting of the NRC Cooperative Severe Accident Research Program (CSARP-92) held at the Hyatt Regency Hotel in Bethesda, Maryland, May 4-8, 1992. The report is taken mainly from coverage of the meeting published in the June 5, 1992, issue of Atomic Energy Clearinghouse. Results of this meeting are formalized at the Water Reactor Safety Information Meetings (WRSIM) that are held annually in October. Nuclear Safety summarizes the annual WRSIM meetings and provides a list of the presentations that were given. Interested readers are encouraged to review listed topics to identify specific topic areas in severe accident research. Sessions were held on in-vessel core melt progression; fuel-coolant interactions; fission-product behavior; direct containment heating; and severe accident code development, assessment, and validation. Summaries of the individual technical sessions and the current state of the art in these areas were given by the chairmen

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

Statistical modelling of the frequency and severity of road accidents

DEFF Research Database (Denmark)

Janstrup, Kira Hyldekær

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

Use of PSA and severe accident assessment results for the accident management

International Nuclear Information System (INIS)

Jang, S. H.; Kim, H. G.; Jang, H. S.; Moon, S. K.; Park, J. U.

1993-12-01

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management

Use of PSA and severe accident assessment results for the accident management

Energy Technology Data Exchange (ETDEWEB)

Jang, S H; Kim, H G; Jang, H S; Moon, S K; Park, J U [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

1993-12-15

The objectives for this study are to investigate the basic principle or methodology which is applicable to accident management, by using the results of PSA and severe accident research, and also facilitate the preparation of accidents management program in the future. This study was performed as follows: derivation of measures for core damage prevention, derivation of measures for accident mitigation, application of computerized tool to assess severe accident management.

Profiling of high-frequency accident locations by use of association rules

OpenAIRE

GEURTS, Karolien; WETS, Geert; BRIJS, Tom; VANHOOF, Koen

2003-01-01

In Belgium, traffic safety is one of the government's highest priorities. The identification and profiling of black spots and black zones (geographical locations with high concentrations of traffic accidents) in terms of accident-related data and location characteristics must provide new insights into the complexity and causes of road accidents, which, in turn, provide valuable input for governmental actions. Association rules were used to identify accident-related circumstances that frequent...

A study on the estimation method of nuclear accident risk cost

International Nuclear Information System (INIS)

Matsuo, Yuji

2016-01-01

The methodology of estimating nuclear accident risk cost, as a part of nuclear power generation cost, has hardly been established due mainly to the extremely wide range of the estimation of the accident frequency. This study estimates the expected nuclear accident frequency for Japan, making use of the method of Bayesian statistics, which exploits both the information obtained by Probabilistic Risk Assessment (PRA) and the observed historical accident frequencies. Using the PRA estimation of the Containment Failure Frequency (CFF) for Tomari nuclear power plant unit 3 of Hokkaido Electric Power Company (average: 2.1 x 10 -4 , 95th percentile: 7.7 x 10 -4 ) and the actual large-scale accident frequency (once in 1,460 reactor-years), the posterior CFF was estimated at 3.8 x 10 -4 . This study also took into account the 'external' factor causing unexpected nuclear accidents, concluding that such factor could result in higher CFF estimations, especially with larger observed accident numbers. (author)

Program of long term immunological monitoring of population groups exposed to risk factors due to Chernobyl accident

International Nuclear Information System (INIS)

Petrov, R.V.; Oradovskaya, I.V.; Pinegin, B.V.

1991-01-01

Program for long-term monitoring of health and immune status of people exposed to ionizing radiation during efforts to to eliminate effects of Chernobyl accident and those subjected to radiation factor effect linked with their living within the contanimated territories is considered. Program purpose consists in investigation into non-stochastic and stochastic immunological effects resulting from radiation factor effect under extreme conditions and under prolonged effect of low doses; in acquisition of new data on immune system state during early after the accident and delayed periods in people which took part in recovery efforts during the accident effect elimination and those who live within the controlled territories. Program is directed to early diagnostics of immune-depented pathologic states and other disfunctions of immunity during prenosologic period, and to development and conduction of combined preventive and treatment-and-sanitary measures using immunomodulation

Radiological accidents potentially important to human health risk in the U.S. Department of Energy waste management program

International Nuclear Information System (INIS)

Mueller, C.; Roglans-Ribas, J.; Folga, S.; Nabelssi, B.; Jackson, R.

1995-01-01

Human health risks as a consequence of potential radiological releases resulting from plausible accident scenarios constitute an important consideration in the US Department of Energy (DOE) national program to manage the treatment, storage, and disposal of wastes. As part of this program, the Office of Environmental Management (EM) is currently preparing a Programmatic Environmental Impact Statement (PEIS) that evaluates the risks that could result from managing five different waste types. This paper (1) briefly reviews the overall approach used to assess process and facility accidents for the EM PEIS; (2) summarizes the key inventory, storage, and treatment characteristics of the various DOE waste types important to the selection of accidents; (3) discusses in detail the key assumptions in modeling risk-dominant accidents; and (4) relates comparative source term results and sensitivities

Experimental programs and facilities for ASTRID development related to the Severe Accident Issue

International Nuclear Information System (INIS)

Journeau, C.; Suteau, C.; Trotignon, L.; Willermoz, G.; Ducros, G.; Courouau, J.L.; Ruggieri, J.M.; Serre, F.

2013-01-01

A comprehensive experimental program has been launched in order to gain new data in support of the severe accident studies related to the ASTRID demonstrator. The main new issues with respect to the historic experimental database are mainly related to new design options: heterogeneous core with thick pins; new materials; new severe accident mitigation systems such as - corium discharge channels; - core-catcher with sacrificial materials; - some issues remaining open as Fuel Coolant Interaction. Experiments are needed both in-pile and out of pile: - Depending on the objectives, the out of pile experiments can be conducted - with simulant; - with prototypic corium; - or with irradiated fuel. A new large scale corium facility, FOURNAISE, must be built to fulfill this program. Already, experimental R&D started in existing facilities, such as VITI or CORRONA

Severe accident management. Prevention and Mitigation

International Nuclear Information System (INIS)

1992-01-01

Effective planning for the management of severe accidents at nuclear power plants can produce both a reduction in the frequency of such accidents as well as the ability to mitigate their consequences if and when they should occur. This report provides an overview of accident management activities in OECD countries. It also presents the conclusions of a group of international experts regarding the development of accident management methods, the integration of accident management planning into reactor operations, and the benefits of accident management

Improving aircraft accident forecasting for an integrated plutonium storage facility

International Nuclear Information System (INIS)

Rock, J.C.; Kiffe, J.; McNerney, M.T.; Turen, T.A.

1998-06-01

Aircraft accidents pose a quantifiable threat to facilities used to store and process surplus weapon-grade plutonium. The Department of Energy (DOE) recently published its first aircraft accident analysis guidelines: Accident Analysis for Aircraft Crash into Hazardous Facilities. This document establishes a hierarchy of procedures for estimating the small annual frequency for aircraft accidents that impact Pantex facilities and the even smaller frequency of hazardous material released to the environment. The standard establishes a screening threshold of 10 -6 impacts per year; if the initial estimate of impact frequency for a facility is below this level, no further analysis is required. The Pantex Site-Wide Environmental Impact Statement (SWEIS) calculates the aircraft impact frequency to be above this screening level. The DOE Standard encourages more detailed analyses in such cases. This report presents three refinements, namely, removing retired small military aircraft from the accident rate database, correcting the conversion factor from military accident rates (accidents per 100,000 hours) to the rates used in the DOE model (accidents per flight phase), and adjusting the conditional probability of impact for general aviation to more accurately reflect pilot training and local conditions. This report documents a halving of the predicted frequency of an aircraft impact at Pantex and points toward further reductions

Effects of the Chernobyl accident on public perceptions of nuclear plant accident risks

International Nuclear Information System (INIS)

Lindell, M.K.; Perry, R.W.

1990-01-01

Assessments of public perceptions of the characteristics of a nuclear power plant accident and affective responses to its likelihood were conducted 5 months before and 1 month after the Chernobyl accident. Analyses of data from 69 residents of southwestern Washington showed significant test-retest correlations for only 10 of 18 variables--accident likelihood, three measures of impact characteristics, three measures of affective reactions, and hazard knowledge by governmental sources. Of these variables, only two had significant changes in mean ratings; frequency of thought and frequency of discussion about a nearby nuclear power plant both increased. While there were significant changes only for two personal consequences (expectations of cancer and genetic effects), both of these decreased. The results of this study indicate that more attention should be given to assessing the stability of risk perceptions over time. Moreover, the data demonstrate that experience with a major accident can actually decrease rather than increase perceptions of threat

Pilot program: NRC severe reactor accident incident response training manual: Severe reactor accident overview

International Nuclear Information System (INIS)

McKenna, T.J.; Martin, J.A.; Miller, C.W.; Hively, L.M.; Sharpe, R.W.; Giitter, J.G.; Watkins, R.M.

1987-02-01

This pilot training manual has been written to fill the need for a general text on NRC response to reactor accidents. The manual is intended to be the foundation for a course for all NRC response personnel. Severe Reactor Accident Overview is the second in a series of volumes that collectively summarize the US Nuclear Regulatory Commission (NRC) emergency response during severe power reactor accidents and provide necessary background information. This volume describes elementary perspectives on severe accidents and accident assesment. Each volume serves, respectively, as the text for a course of instruction in a series of courses. Each volume is accompanied by an appendix of slides that can be used to present this material. The slides are called out in the text

TMI-2 accident evaluation program sample acquisition and examination plan. Executive summary

International Nuclear Information System (INIS)

Russell, M.L.; McCardell, R.K.; Broughton, J.M.

1985-12-01

The purpose of the TMI-2 Accident Evaluation Program Sample Acquisition and Examination (TMI-2 AEP SA and E) program is to develop and implement a test and inspection plan that completes the current-condition characterization of (a) the TMI-2 equipment that may have been damaged by the core damage events and (b) the TMI-2 core fission product inventory. The characterization program includes both sample acquisitions and examinations and in-situ measurements. Fission product characterization involves locating the fission products as well as determining their chemical form and determining material association

Proceedings of the Second NASA Aviation Safety Program Weather Accident Prevention Review

Science.gov (United States)

Martzaklis, K. Gus (Compiler)

2003-01-01

The Second NASA Aviation Safety Program (AvSP) Weather Accident Prevention (WxAP) Annual Project Review held June 5-7, 2001, in Cleveland, Ohio, presented the NASA technical plans and accomplishments to the aviation community. NASA-developed technologies presented included an Aviation Weather Information System with associated digital communications links, electronic atmospheric reporting technologies, forward-looking turbulence warning systems, and turbulence mitigation procedures. The meeting provided feedback and insight from the aviation community of diverse backgrounds and assisted NASA in steering its plans in the direction needed to meet the national safety goal of 80-percent reduction of aircraft accidents by 2007. The proceedings of the review are enclosed.

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

Energy Technology Data Exchange (ETDEWEB)

Fleming, K.N., E-mail: KarlFleming@comcast.net [KNF Consulting LLC, Spokane, WA (United States); Lydell, B.O.Y. [SIGMA-PHASE INC., Vail, AZ (United States)

2016-08-15

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

Insights into location dependent loss-of-coolant-accident (LOCA) frequency assessment for GSI-191 risk-informed applications

International Nuclear Information System (INIS)

Fleming, K.N.; Lydell, B.O.Y.

2016-01-01

Highlights: • Role of operating experience in loss-of-coolant-accident (LOCA) frequency assessment. • Plant-to-plant variability in calculated LOCA frequency. • Frequency of double-ended-guillotine-break (DEGB). • Uncertainties in LOCA frequencies. • Risk management insights. - Abstract: As a tribute to the published work by S.H. Bush, S. Beliczey and H. Schulz, this paper assesses the progress with methods and techniques for quantifying the reliability of piping systems in commercial nuclear power plants on the basis of failure rate estimates derived from field experience data in combination with insights and results from probabilistic fracture mechanics analyses and expert elicitation exercises. This status assessment is made from a technical perspective obtained through development of location-specific loss-of-coolant-accident (LOCA) frequencies for input to risk-informed resolution of the generic safety issue (GSI) 191. The methods and techniques on which these GSI-191 applications are based build on a body of work developed by the authors during a period spanning more than two decades. The insights that are presented and discussed in this paper cover today’s knowledge base concerning how to utilize a risk-informed approach to the assessment of piping reliability in the context of probabilistic risk assessment (PRA) in general and the resolution of GSI-191 in particular. Specifically the paper addresses the extent to which LOCA frequencies vary from location to location within a reactor coolant system pressure boundary (RCPB) for a given plant as well as vary from plant to plant, and the reasons for these variabilities. Furthermore, the paper provides the authors’ perspectives on interpretations and applications of information extracted from an expert elicitation process to obtain LOCA frequencies as documented in NUREG-1829 and how to apply this information to GSI-191. Finally, this paper documents technical insights relative to mitigation of

Accidents in family forestry's firewood production.

Science.gov (United States)

Lindroos, Ola; Aspman, Emma Wilhelmson; Lidestav, Gun; Neely, Gregory

2008-05-01

Firewood is commonly used around the world, but little is known about the work involved in its production and associated accidents. The objectives were to identify relationships between accidents and time exposure, workers' age and sex, equipment used and work activities in family forestry's firewood production. Data from a postal survey in Northern Sweden were compared to a database of injuries in the same region. Most accidents occurred to 50-69 year old men, who also worked most hours. No significant differences in sex and age were found between expected and recorded accident frequencies when calculated from total work hours; however, when calculated using numbers of active persons significant differences were found for both age and sex. Frequency of accidents per unit worked time was higher for machine involving activities than for other activities. Accidents that occurred when using wedge splitter machines were responsible for most of this overrepresentation. Fingers were the most commonly injured body parts. Mean accident rate for the equipment used was 87 accidents per million work hours, and the rate was highest for wedge splitters (122 accidents per million work hours). Exposure to elevated risks due to violation of safety procedures is discussed, as well as possible preventative measures.

Frequency of work zone accidents on construction projects : final report.

Science.gov (United States)

2005-08-01

The overall objective of this research was to study work zone accidents in New York State, with particular attention to the : occurrence and mitigation of rear-end vehicle accidents. The specific objectives were to: : - Recommend changes to the NYSDO...

Facility accident considerations in the US Department of Energy Waste Management Program

International Nuclear Information System (INIS)

Mueller, C.

1994-01-01

A principal consideration in developing waste management strategies is the relative importance of Potential radiological and hazardous releases to the environment during postulated facility accidents with respect to protection of human health and the environment. The Office of Environmental Management (EM) within the US Department of Energy (DOE) is currently formulating an integrated national program to manage the treatment, storage, and disposal of existing and future wastes at DOE sites. As part of this process, a Programmatic Environmental impact Statement (PEIS) is being prepared to evaluate different waste management alternatives. This paper reviews analyses that have been Performed to characterize, screen, and develop source terms for accidents that may occur in facilities used to store and treat the waste streams considered in these alternatives. Preliminary results of these analyses are discussed with respect to the comparative potential for significant releases due to accidents affecting various treatment processes and facility configurations. Key assumptions and sensitivities are described

Regulatory approach to accident management in Sweden

International Nuclear Information System (INIS)

Hoegberg, L.

1989-01-01

The Swedish accident management program includes the following components: definition of overall safety and radiation protection objectives for the program; definition of appropriate accident management strategies to reach these objectives, based on plant-specific severe accident analysis; development and installation of appropriate accident management systems and associated management procedure; definition of roles and resposibilities for plant staff involved in accident management and implementation of appropriate training programs. The discussion of these components tries to highlight the basic technical concepts and approaches and the underlying safety philosophy rather than going into design details. 5 figs., 7 refs

Radioprotection program to attend of radiation accidents

International Nuclear Information System (INIS)

Fajardo, P.W.; Costa Silva, L.H. da; Rosa, R.

1989-04-01

The aspects of a radioprotection program to be implanted in hospitals to cases of medical treatment to external and internal contamined people are presented. It is based in the experience acquired in the coordination of radioprotection of the Marcilio Dias Naval Medical Center, Rio de Janeiro, due to accident happened in Goiania in 1987. The infra-structure necessary of a ward and the procedures of radioprotection to acess control, entrance and way out of material area and patients monitoring, decontamination and the activities in the support area such as the control and maintenance of detection equipments and radioactive waste management are described. The radiologic protection materials necessaries and the quantity of radioactive waste generated by patient for day are estimated. (V.R.B.)

Ten years of work in Cuban Program to children from areas affected of Chernobyl accidents

International Nuclear Information System (INIS)

Garcia, Omar; Medina J, Julio

2001-01-01

In April of 2000 the Cuban program for specialised medical attention to children from areas affected by the radioactive contamination from the Chernobyl accident completed 10 years of work. In the program had been assisted more than 18 000 children and adults, among them 50 Brazilians related with the Goiania accident. A group of medical procedures and dosimetric, biomedical and psychological investigations have been carried out in the program. The main significant medical attention activities are the treatment of haematological disorders, among them, 120 leukaemia, the realisation of bone marrow transplants and the treatment of endocrinological and neoplasic illnesses. The dosimetric studies has allowed to create a database that accumulates information about internal contamination for 137Cs, internal, external, and total doses, of 7000 children. The analysis of the behaviour of all the medical information that is generated in the program in function of the contamination of the land and of the internal contamination of the children is also performed. The program has accumulated an experience of interest for physicians, psychologists and radiological emergencies experts.(author)

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

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage

International Nuclear Information System (INIS)

Budnitz, R.J.; Davis, P.R.; Ravindra, M.K.; Tong, W.H.

1994-08-01

In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ''internal initiators.'' This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10 -7 /year

Application of the accident management information needs methodology to a severe accident sequence

International Nuclear Information System (INIS)

Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

1989-01-01

The U.S. Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe accident research results to enhance the capability of plant operating personnel to effectively manage severe accidents. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed as part of the research program designed to resolve this issue. The methodology identifies the information needs of the plant personnel during a wide range of accident conditions, the existing plant measurements capable of supplying these information needs and what, if any minor additions to instrument and display systems would enhance the capability to manage accidents, known limitations on the capability of these measurements to function properly under the conditions that will be present during a wide range of severe accidents, and areas in which the information systems could mislead plant personnel. This paper presents an application of this methodology to a severe accident sequence to demonstrate its use in identifying the information which is available for management of the event. The methodology has been applied to a severe accident sequence in a Pressurized Water Reactor with a large dry containment. An examination of the capability of the existing measurements was then performed to determine whether the information needs can be supplied

Supervisor's accident investigation handbook

International Nuclear Information System (INIS)

1980-02-01

This pamphlet was prepared by the Environmental Health and Safety Department (EH and S) of Lawrence Berkeley Laboratory (LBL) to provide LBL supervisors with a handy reference to LBL's accident investigation program. The publication supplements the Accident and Emergencies section of LBL's Regulations and Procedures Manual, Pub. 201. The present guide discusses only accidents that are to be investigated by the supervisor. These accidents are classified as Type C by the Department of Energy (DOE) and include most occupational injuries and illnesses, government motor-vehicle accidents, and property damages of less than $50,000

Aircraft accident analysis for emergency planning and safety analysis

International Nuclear Information System (INIS)

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

1996-01-01

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

Westinghouse accident tolerant fuel program. Current results and future plans

Energy Technology Data Exchange (ETDEWEB)

Ray, Sumit; Xu, Peng; Lahoda, Edward; Hallstadius, Lars; Boylan, Frank [Westinghouse Electric Company LLC, Hopkins, SC (United States)

2016-07-15

This paper discusses the current status, results from initial tests, as well as the future direction of the Westinghouse's Accident Tolerant Fuel (ATF) program. The current preliminary testing is addressed that is being performed on these samples at the Massachusetts Institute of Technology (MIT) test reactor, initial results from these tests, as well as the technical learning from these test results. In the Westinghouse ATF approach, higher density pellets play a significant role in the development of an integrated fuel system.

Latin American radio pathology net, training program for the medical response in cases of accidents

International Nuclear Information System (INIS)

Perez, M. R.; Valverde, N.; Sanhueza, S.; Di Tramo, J. L.; Gisone, P.; Cardenas, J.

2003-01-01

In Latin America exists a wide application of the nuclear techniques in different fields. These tendencies, to the increasing use of nuclear techniques, should carry out the invigoration of the infrastructures and the development of the authorities in radiation protection charge of the regulate and control the sources and practical, associated to ionizing radiation, guided to guarantee their safe use and in consequence to minimize the derived risks of the same ones. In spite of the efforts before mentioned a potential possibility of occurrence of accidental radiological events linked to human errors and violation's of the principles of the radiation protection exists. Reason that they advise to have response capacities to confront and to mitigate the consequences in situations of radiological accidents, including in the same ones the medical assistance of the accident victims. However, the radiological accidents happened in the international environment in the last decades, they have demonstrated inability paradoxically to confront with effectiveness these fortuitous events. Being characterized additionally by the insufficient training of the medical professionals to interpret and to act in consequence before the prejudicial effects to the health of the ionizing radiation. In our geographical context this situation is even more complicated, if we consider the happened radiological accidents of span that put in risk the life and the health of people involved in the same ones, for examples: Argentina (1968, 1963), Brazil (1985, 1987, 1995), Costa Rica (1996). El Salvador (1989), Mexico (1962, 1983, 1984) and Peru (1984, 1999). These reasons justify the development of an action program with the purpose of the consolidate and to integrate the capacity of response of our countries as regards radiological emergencies. Regrettably in Latin American subsist inequalities the development radiation protection programs, that propitiate the possibility of accidental situations

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1

Energy Technology Data Exchange (ETDEWEB)

Jo, J.; Lin, C.C.; Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)] [and others

1995-05-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6.

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1

International Nuclear Information System (INIS)

Jo, J.; Lin, C.C.; Neymotin, L.

1995-05-01

During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6

Does the prevalence of latent toxoplasmosis and frequency of Rhesus-negative subjects correlate with the nationwide rate of traffic accidents?

Science.gov (United States)

Flegr, Jaroslav; Dama, Madhukar

2014-12-01

Latent toxoplasmosis is probably the most common protistan parasitic disease with many indirect negative impacts on human health. One of the important impacts is impaired psychomotor function leading to reduced driving efficiency in Toxoplasma-seropositive subjects. Numerous case-control studies have established a positive relation between the seroprevalence of Toxoplasma gondii (Nicolle et Manceaux, 1908) and probability of traffic accidents in study populations. The prevalence of toxoplasmosis varies between populations according to local geographical conditions, hygienic practices and kitchen habits. Similarly, we see a striking variation in the incidence of traffic accidents across countries. Hence, we compiled the largest ever data set on the seroprevalence of toxoplasmosis and tried to understand its role in traffic accident-related deaths and disabilities across 87 countries. Simple non-parametric analysis showed a positive and strong relation of T. gondii seroprevalence and traffic accident related disabilities. Further, we conducted multivariate analysis to control for confounding factors. After controlling for wealth, geographical latitude, health of population, length of roads and number of vehicles, the correlation disappeared. When the frequency of RhD negativity and its interaction with toxoplasmosis were included into the model, the effects of toxoplasmosis seemingly returned. However, the analysed data suffered from the problem of multicollinearity. When a proper method of analysis, ridge regression, was applied, the effects of toxoplasmosis prevalence and RhD negativity frequency disappeared again. The existence of a strong correlation between the prevalence of toxoplasmosis and health of population in particular countries, which was the probable cause of multicollinearity and possible reason for the negative result of the present study, suggests that 'asymptomatic' latent toxoplasmosis could have a large impact on public health.

Return to driving after severe traumatic brain injury: increased risk of traffic accidents and personal responsibility.

Science.gov (United States)

Bivona, Umberto; DʼIppolito, Mariagrazia; Giustini, Marco; Vignally, Pascal; Longo, Eloise; Taggi, Franco; Formisano, Rita

2012-01-01

To determine the frequency of road traffic accidents among individuals who start or resume driving after severe traumatic brain injury (TBI) and to investigate their responsibility for these accidents. Observational/retrospective study. Sixty adults with severe TBI and their caregivers. Return to Driving Questionnaire and Glasgow Outcome Scale. Thirty of the 60 participants started to drive or resumed driving after TBI. Nineteen (63%) of them were involved in traffic accidents, with personal responsibility in 26 of 36 after return to driving. Participants caused a significantly higher number of accidents after TBI than before. The ability to drive is frequently compromised after severe TBI. Specific rehabilitation of this complex activity should be a main goal of social reintegration programs in this population.

[Development and effect analysis of web-based instruction program to prevent elementary school students from safety accidents].

Science.gov (United States)

Chung, Eun-Soon; Jeong, Ihn-Sook; Song, Mi-Gyoung

2004-06-01

This study was aimed to develop a WBI(Web Based Instruction) program on safety for 3rd grade elementary school students and to test the effects of it. The WBI program was developed using Macromedia flash MX, Adobe Illustrator 10.0 and Adobe Photoshop 7.0. The web site was http://www.safeschool.co.kr. The effect of it was tested from Mar 24, to Apr 30, 2003. The subjects were 144 students enrolled in the 3rd grade of an elementary school in Gyungju. The experimental group received the WBI program lessons while each control group received textbook-based lessons with visual presenters and maps, 3 times. Data was analyzed with descriptive statistics, and chi2 test, t-test, and repeated measure ANOVA. First, the WBI group reported a longer effect on knowledge and practice of accident prevention than the textbook-based lessons, indicating that the WBI is more effective. Second, the WBI group was better motivated to learn the accident prevention lessons, showing that the WBI is effective. As a result, the WBI group had total longer effects on knowledge, practice and motivation of accident prevention than the textbook-based instruction. We recommend that this WBI program be used in each class to provide more effective safety instruction in elementary schools.

Frequency of Specific Categories of Aviation Accidents and Incidents During 2001-2010

Science.gov (United States)

Evans, Joni K.

2014-01-01

The purpose of this study was to determine the types of accidents or incidents that are most important to the aviation safety risk. All accidents and incidents from 2001-2010 were assigned occurrence categories based on the taxonomy developed by the Commercial Aviation Safety Team/International Civil Aviation Organization (CAST/ICAO) Common Taxonomy Team (CICTT). The most frequently recorded categories were selected within each of five metrics: total accidents, fatal accidents, total injuries, fatal injuries and total incidents. This analysis was done separately for events within Part 121, Scheduled Part 135, Non-Scheduled Part 135 and Part 91. Combining those five sets of categories resulted in groups of between seven and eleven occurrence categories, depending on the flight operation. These groups represent 65-85% of all accidents and 68-81% of incidents.

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...

Pattern extraction for high-risk accidents in the construction industry: a data-mining approach.

Science.gov (United States)

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

2016-09-01

Accidents involving falls and falling objects (group I) are highly frequent accidents in the construction industry. While being hit by a vehicle, electric shock, collapse in the excavation and fire or explosion accidents (group II) are much less frequent, they make up a considerable proportion of severe accidents. In this study, multiple-correspondence analysis, decision tree, ensembles of decision tree and association rules methods are employed to analyse a database of construction accidents throughout Iran between 2007 and 2011. The findings indicate that in group I, there is a significant correspondence among these variables: time of accident, place of accident, body part affected, final consequence of accident and lost workdays. Moreover, the frequency of accidents in the night shift is less than others, and the frequency of injury to the head, back, spine and limbs are more. In group II, the variables time of accident and body part affected are mostly related and the frequency of accidents among married and older workers is more than single and young workers. There was a higher frequency in the evening, night shifts and weekends. The results of this study are totally in line with the previous research.

Thermal-hydraulic uncertainties affecting severe accident progression

International Nuclear Information System (INIS)

Haskin, F.E.; Behr, V.L.

1984-01-01

To provide the proper technical bases for decisions regarding severe accidents, the US Nuclear Regulatory Commission (NRC) is sponsoring the following activities: (a) a variety of severe accident research programs, combined under the Severe Accident Research Plan; (b) nationwide task forces on containment loading, containment response, and fission product source terms; (c) a review by the American Physical Society of state-of-the-art methods for calculating radiological source terms; and (d) technical exchange meetings with the Industry Degraded Core (IDCOR) program. One of the means for integrating this developing array of technical information is the Severe Accident Risk Reduction Program (SARRP). One of the current SARRP objectives is to utilize insights gained from the activities listed above to characterize the relative likelihoods of competing containment failure modes for core-melt accidents

Presentation of the ressac research program (rehabilitation of soils and surfaces after an accident)

International Nuclear Information System (INIS)

L'Homme, A.

1989-11-01

If, despite all the precautions taken in nuclear power plants, a severe accident was to occur in France involving extensive release of radioactive materials to the environment, existing emergency plans would be implemented enabling urgent decisions to be made with regard to the immediate protection of the population: confinement indoors, evacuation, distribution of stable iodine, etc. But, at a later stage, mean and long term actions would have to be carried out to decontaminate the polluted areas and limit subsequent contamination of the food chain, with a view to enabling the populations concerned to return to normal life. These actions would concern, in decreasing order of priority and using the WHO and IAEA definitions, the near field, closest to the accident site, and the far field, subjected to the direct impact of fallout. They should be aimed at reducing external exposure due to deposition and internal exposure by inhalation of radioactive products re-suspended in the atmosphere and by ingestion of products for human consumption. In the context of IPSN (Institute of Protection and Nuclear Safety) research and development programs on severe accidents, the RESSAC program was defined in 1985 for the purpose of studying methods and means of rehabilitating the near field and controlling problems related to the far field. Elaboration of the program is presently proceeding at the Nuclear Research Center of CADARACHE, focussed on the following main topics: assessment of what happens to the radionuclides deposited on the soil and vegetation, determination of priorities and how to intervene, management of the waste produced

Presentation of the RESSAC research program (REhabilitation of Soils and Surfaces after an ACcident)

International Nuclear Information System (INIS)

L'Homme, A.; Parmentier, N.; Legrand, B.; Fache, P.

1992-01-01

If, despite all the precautions taken in nuclear power plants, a severe accident were to occur in France involving extensive release of radioactive materials to the environment, existing emergency plans would be implemented enabling urgent decisions to be made with regard to the immediate protection of the population: confinement indoors, evacuation, distribution of stable iodine, etc. But, at a later stage, mean and long term actions would have to be carried out to decontaminate the polluted areas and limit subsequent contamination of the food chain, with a view to enabling the populations concerned to return to normal life. These actions would concern, in decreasing order of priority and using the WHO and IAEA definitions, the near field, closest to the accident site, and the far field, subjected to the direct impact of fallout. They should be aimed at reducing external exposure due to deposition and internal exposure by inhalation of radioactive products re-suspended in the atmosphere and by ingestion of products for human consumption. In the context of IPSN research and development programs on severe accidents, the RESSAC program was defined in 1985 for the purpose of studying methods and means of rehabilitating the near field and controlling problems related to the far field. Elaboration of the program is presently proceeding at the Nuclear Research Center of Cadarache, focussed on the following main topics: assessment of what happens to the radionuclides deposited on the soil and vegetation, determination of priorities and how to intervene, management of the waste produced. (author). 4 refs

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit-1: Analysis of core damage frequency from internal events during mid-loop operations. Appendices F-H, Volume 2, Part 4

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Bley, D.; Johnson, D.; Holmes, B.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

Part-time farmers and accidents with agricultural machinery: a moderated mediated model on the role played by frequency of use and unsafe beliefs.

Science.gov (United States)

Caffaro, Federica; Roccato, Michele; Micheletti Cremasco, Margherita; Cavallo, Eugenio

2018-01-25

We aimed at testing a model of the direct and indirect effects of being a part-time farmer on the probability of being involved in an agricultural machinery-related accident, considering the role played by unsafe beliefs and the frequency of use of machinery. Two-hundred and fifty-two Italian men, regular users of agricultural machinery (age: Mean = 45.1 years, standard Deviation = 17.5), were administered a paper-and-pencil questionnaire addressing their relation with work, unsafe beliefs, and previous experience of machinery-related accidents. Being a part-time farmer showed a positive association with unsafe beliefs only among occasional machinery users. Unsafe beliefs in turn showed a positive association with accidents. The study gave a novel contribution to the knowledge of the chain of events connecting part-time farmers with machinery-related accidents. Preventive training interventions targeting part-timer farmers using agricultural machinery just occasionally should be developed.

Frequency of dicentrics in Ukrainian children and teenagers from areas near Chernobyl 20 years after the nuclear power plant accident

International Nuclear Information System (INIS)

Montoro, A.; Sebastia, N.; Candela, C.; Francesc Barquinero, J.; Soriano, J. M.; Almonacid, M.; Sahuquillo, V.; Alonso, O.; Cervera, J.; Such, E.; Silla, M. A.; Arnal, C.; Villaescusa, J. I.

2012-01-01

In this work, frequency of dicentrics in the peripheral blood of 55 children and teenagers from Ukrainian areas affected by the Chernobyl accident are studied, in order to assess the possible existence of chromosomal damage due to radiation, and to estimate absorbed dose through biological dosimetry. A total of 36 dicentric chromosomes found in 53477 analyzed cells reflected a low dicentric frequency (below the baseline limit). From these results, it can be concluded that, within the detection limits of the used technique, no overexposure to radiation was detected in these children. (Author) 18 refs.

Approach to accident management in RBMK-1500

International Nuclear Information System (INIS)

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

2008-01-01

In order to ensure the safe operation of the nuclear power plants accident management programs are being developed around the world. These accident management programs cover the whole spectrum of accidents, including severe accidents. A lot of work is done to investigate the severe accident phenomena and implement severe accident management in NPPs with vessel-type reactors, while less attention is paid to channel-type reactors CANDU and RBMK. Ignalina NPP with RBMK-1500 reactor has implemented symptom based emergency operation procedures, which cover management of accidents until the core damage and do not extend to core damage region. In order to ensure coverage of the whole spectrum of accidents and meet the requirements of IAEA the severe accident management guidelines have to be developed. This paper presents the basic principles and approach to management of beyond design basis accidents at Ignalina NPP. In general, this approach could be applied to NPPs with RBMK-1000 reactors that are available in Russia, but the design differences should be taken into account

Release of fission products during controlled loss-of-coolant accidents and hypothetical core meltdown accidents

International Nuclear Information System (INIS)

Albrecht, H.; Malinauskas, A.P.

1978-01-01

A few years ago the Projekt Nukleare Sicherheit joined the United States Nuclear Regulatory Commission in the development of a research program which was designed to investigate fission product release from light water reactor fuel under conditions ranging from spent fuel shipping cask accidents to core meltdown accidents. Three laboratories have been involved in this cooperative effort. At Argonne National Laboratory (ANL), the research effort has focused on noble gas fission product release, whereas at Oak Ridge National Laboratory (ORNL) and at Kernforschungszentrum Karlsruhe (KfK), the studies have emphasized the release of species other than the noble gases. In addition, the ORNL program has been directed toward the development of fission product source terms applicable to analyses of spent fuel shipping cask accidents and controlled loss-of-coolant accidents, and the KfK program has been aimed at providing similar source terms which are characteristic of core meltdown accidents. The ORNL results are presented for fission product release from defected fuel rods into a steam atmosphere over the temperature range 500 to 1200 0 C, and the KfK results for release during core meltdown sequences

Reactor materials program process water component failure probability

International Nuclear Information System (INIS)

Daugherty, W. L.

1988-01-01

The maximum rate loss of coolant accident for the Savannah River Production Reactors is presently specified as the abrupt double-ended guillotine break (DEGB) of a large process water pipe. This accident is not considered credible in light of the low applied stresses and the inherent ductility of the piping materials. The Reactor Materials Program was initiated to provide the technical basis for an alternate, credible maximum rate LOCA. The major thrust of this program is to develop an alternate worst case accident scenario by deterministic means. In addition, the probability of a DEGB is also being determined; to show that in addition to being mechanistically incredible, it is also highly improbable. The probability of a DEGB of the process water piping is evaluated in two parts: failure by direct means, and indirectly-induced failure. These two areas have been discussed in other reports. In addition, the frequency of a large bread (equivalent to a DEGB) in other process water system components is assessed. This report reviews the large break frequency for each component as well as the overall large break frequency for the reactor system

The experience gives the Cuban program with children gives territories affected by the Chernobyl accident

International Nuclear Information System (INIS)

Garcia, O.; Llanes, R.

1998-01-01

From 1990 it works in Cuba a program destined to offer medical attention you specialize and to develop a plan sanatoria gives rehabilitation with children provided the different areas affected by the contamination radioactive resultant to the Chernobyl accident

EPRI research on accident management

International Nuclear Information System (INIS)

Oehlberg, R.N.; Chao, J.

1991-01-01

The paper discusses Nuclear Regulatory Commission (NRC) efforts regarding severe reactor accident management and the Nuclear Management and Resources Council (NUMAEX), activities. (EPRI) Electric Power Research Institute accident management program consists of the two products just mentioned plus one related to severe accident plant status information and the MAAP 4.0 computer code. These are briefly discussed

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-01-31

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

Chromosome aberrations in Norwegian reindeer following the Chernobyl accident

International Nuclear Information System (INIS)

Røed, K.H.; Jacobsen, M.

1995-01-01

Chromosome analyses were carried out on peripheral blood lymphocytes of semi-domestic reindeer in Norway which had been exposed to varying amounts of radiocesium emanating from the Chernobyl accident. The sampling was done in the period 1987-1990. The material included 192 reindeer, originating from four herds in central Norway, an area considerably affected by fallout from the Chernobyl accident, and from three herds in northern Norway which was unaffected by fallout from the accident. Significant heterogeneity in the distribution of chromosome aberrations between herds was observed. The pattern of chromosome aberration frequencies between herds was not related to the variation in radiocesium exposure from the Chernobyl accident. Other factors than the Chernobyl accident appear therefore to be of importance for the distribution of aberration frequencies found among present herds. Within the most contaminated area the reindeer born in 1986 showed significantly more chromosome aberrations than those born both before and after 1986. This could suggest that the Chernobyl accident fallout created an effect particularly among calves, during the immediate post-accident period in the most exposed areas

Station blackout transient at the Browns Ferry Unit 1 Plant: a severe accident sequence analysis (SASA) program study

International Nuclear Information System (INIS)

Schultz, R.R.

1982-01-01

Operating plant transients are of great interest for many reasons, not the least of which is the potential for a mild transient to degenerate to a severe transient yielding core damage. Using the Browns Ferry (BF) Unit-1 plant as a basis of study, the station blackout sequence was investigated by the Severe Accident Sequence Analysis (SASA) Program in support of the Nuclear Regulatory Commission's Unresolved Safety Issue A-44: Station Blackout. A station blackout transient occurs when the plant's AC power from a comemrcial power grid is lost and cannot be restored by the diesel generators. Under normal operating conditions, f a loss of offsite power (LOSP) occurs [i.e., a complete severance of the BF plants from the Tennessee Valley Authority (TVA) power grid], the eight diesel generators at the three BF units would quickly start and power the emergency AC buses. Of the eight diesel generators, only six are needed to safely shut down all three units. Examination of BF-specific data show that LOSP frequency is low at Unit 1. The station blackout frequency is even lower (5.7 x 10 - 4 events per year) and hinges on whether the diesel generators start. The frequency of diesel generator failure is dictated in large measure by the emergency equipment cooling water (EECW) system that cools the diesel generators

Defense Programs Transportation Risk Assessment

International Nuclear Information System (INIS)

Clauss, D.B.

1994-01-01

This paper provides an overview of the methodology used in a probabilistic transportation risk assessment conducted to assess the probabilities and consequences of inadvertent dispersal of radioactive materials arising from severe transportation accidents. The model was developed for the Defense Program Transportation Risk Assessment (DPTRA) study. The analysis incorporates several enhancements relative to previous risk assessments of hazardous materials transportation including newly-developed statistics on the frequencies and severities of tractor semitrailer accidents and detailed route characterization using the 1990 Census data

The impact of vessel speed reduction on port accidents.

Science.gov (United States)

Chang, Young-Tae; Park, Hyosoo

2016-03-19

Reduced-speed zones (RSZs) have been designated across the world to control emissions from ships and prevent mammal strikes. While some studies have examined the effectiveness of speed reduction on emissions and mammal preservation, few have analyzed the effects of reduced ship speed on vessel safety. Those few studies have not yet measured the relationship between vessel speed and accidents by using real accident data. To fill this gap in the literature, this study estimates the impact of vessel speed reduction on vessel damages, casualties and frequency of vessel accidents. Accidents in RSZ ports were compared to non-RSZ ports by using U.S. Coast Guard data to capture the speed reduction effects. The results show that speed reduction influenced accident frequency as a result of two factors, the fuel price and the RSZ designation. Every $10 increase in the fuel price led to a 10.3% decrease in the number of accidents, and the RSZ designation reduced vessel accidents by 47.9%. However, the results do not clarify the exact impact of speed reduction on accident casualty. Copyright © 2016 Elsevier Ltd. All rights reserved.

Application of the accident management information needs methodology to a severe accident sequence

International Nuclear Information System (INIS)

Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E.

1989-01-01

The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel

Application of the accident management information needs methodology to a severe accident sequence

Energy Technology Data Exchange (ETDEWEB)

Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E. (Nuclear Regulatory Commission, Washington, DC (USA))

1989-11-01

The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel.

Regulatory perspective on accident management issues

International Nuclear Information System (INIS)

Barrett, R.J.

1988-01-01

Effective response to reactor accidents requires a combination of emergency operations, technical support and emergency response. The NRC and industry have actively pursued programs to assure the adequacy of emergency operations and emergency response. These programs will continue to receive high priority. By contrast, the technical support function has received relatively little attention from NRC and the industry. The results from numerous PRA studies and the severe accident programs of NRC and the industry have yielded a wealth of insights on prevention and mitigation of severe accidents. The NRC intends to work with the industry to make these insights available to the technical support staffs through a combination of guidance, training and periodic drills

Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

International Nuclear Information System (INIS)

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

2015-01-01

This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

Outline of the Desktop Severe Accident Graphic Simulator Module for OPR-1000

Energy Technology Data Exchange (ETDEWEB)

Park, S. Y.; Ahn, K. I. [KAERI, Daejeon (Korea, Republic of)

2015-05-15

This paper introduce the desktop severe accident graphic simulator module (VMAAP) which is a window-based severe accident simulator using MAAP as its engine. The VMAAP is one of the submodules in SAMEX system (Severe Accident Management Support Expert System) which is a decision support system for use in a severe accident management following an incident at a nuclear power plant. The SAMEX system consists of four major modules as sub-systems: (a) Severe accident risk data base module (SARDB): stores the data of integrated severe accident analysis code results like MAAP and MELCOR for hundreds of high frequency scenarios for the reference plant; (b) Risk-informed severe accident risk data base management module (RI-SARD): provides a platform to identify the initiating event, determine plant status and equipment availability, diagnoses the status of the reactor core, reactor vessel and containment building, and predicts the plant behaviors; (c) Severe accident management simulator module (VMAAP): runs the MAAP4 code with user friendly graphic interface for input deck and output display; (d) On-line severe accident management guidance module (On-line SAMG); provides available accident management strategies with an electronic format. The role of VMAAP in SAMEX can be described as followings. SARDB contains the most of high frequency scenarios based on a level 2 probabilistic safety analysis. Therefore, there is good chance that a real accident sequence is similar to one of the data base cases. In such a case, RI-SARD can predict an accident progression by a scenario-base or symptom-base search depends on the available plant parameter information. Nevertheless, there still may be deviations or variations between the actual scenario and the data base scenario. The deviations can be decreased by using a real-time graphic accident simulator, VMAAP.. VMAAP is a MAAP4-based severe accident simulation model for OPR-1000 plant. It can simulate spectrum of physical processes

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

International Nuclear Information System (INIS)

Hsii, Y.H.

1975-01-01

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

Overview of severe accident research at JAERI

International Nuclear Information System (INIS)

Sugimoto, Jun

1999-01-01

Severe accident research at JAERI aims at the confirmation of the safety margin, the quantification of the associated risk, and the evaluation of the effectiveness of the accident management measures of the nuclear power reactors, in accordance with the government five-year nuclear safety research program. JAERI has been conducting a wide range of severe accident research activities both in experiment and analysis, such as melt coolant interactions, fission product behaviors in coolant system, containment integrity and assessment of accident management measures. Molten core/coolant interaction and in-vessel molten coolability have been investigated in ALPHA Program. MUSE experiments in ALPHA Program has been conducted for the precise energy measurement due to steam explosion in melt jet and stratified geometries. In VEGA Program, which aims at FP release from irradiated fuels at high temperature and high pressure under various atmospheric conditions, the facility construction is almost completed. In WIND Program the revaporization of aerosols due to decay heating and also the integrity of the piping from this heat source are being investigated. Code development activities are in progress for an integrated source term analysis with THALES, fission product behaviors with ART, steam explosion with JASMINE, and in-vessel debris behaviors with CAMP. The experimental analyses and reactor application have made progress by participating international standard problem and code comparison exercises, along with the use of introduced codes, such as SCDAP/RELAP5 and MELCOR. The outcome of the severe accident research will be utilized for the evaluation of more reliable severe accident scenarios, detailed implementation of the accident management measures, and also for the future reactor development, basically through the sophisticated use of verified analytical tools. (author)

Severe accidents and operator training - discussion of potential issues

International Nuclear Information System (INIS)

Vidard, Michel

1997-01-01

R and D programs developed throughout the world allowed significant progress in the understanding of physical phenomena and Severe Accident Management (SAM) programs started in many OECD countries. Basically, the common denominator to all these SAM programs was to provide utility operators with procedures or guidelines allowing to deal with complex situations not formally considered in the Design Basis, including accidents where a significant portion of the core had molten. These SAM procedures or guidelines complement the traditional accident management procedures (event, symptom or physical-state oriented) and should allow operators to deal with a reasonably bounding set of situations. Dealing with operator or crisis team training, it was recognized that training would be beneficial but that training programs were lagging, i.e. though training sessions were either organized or contemplated after implementation of SAM programs, they seemed to be somewhat different from more traditional training sessions on Accident Management. After some explanations on the differences between Design Basis Accidents (DBAs) and Beyond Design Basis Accidents (BDBAs), this paper underlines some potential difficulties for training operators and discuss problems to be addressed by organisms contemplating SAM training sessions consistent with similar activities for less complex events

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

International Nuclear Information System (INIS)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan

Review of literature on the TMI accident and correlation to the LWR Safety Technology Program

Energy Technology Data Exchange (ETDEWEB)

Miller, W.J.

1980-05-01

This report is the result of approximately two man-months of effort devoted to assimilating and comprehending significant publicly available material related to Three Mile Island Unit 2 and events during and subsequent to the accident experienced on March 28, 1979. Those events were then correlated with the Preliminary LWR Safety Technology Program Plan (Preliminary Program Plan) prepared for the US Department of Energy by Sandia National Lab. This report is being submitted simultaneously with the SAI report entitled Preliminary Prioritization of Tasks in the Draft LWR Safety Technology Program Plan.

Frequencies of chromosomal inversions in Drosophila melanogaster in Fukushima after the nuclear power plant accident.

Science.gov (United States)

Itoh, Masanobu; Kajihara, Ryutaro; Kato, Yasuko; Takano-Shimizu, Toshiyuki; Inoue, Yutaka

2018-01-01

In order to investigate genetic impact of a large amount of radionuclides released by the Fukushima Dai-ichi Nuclear Power Plant accident in 2011, we surveyed 2,304 haploid genomes of Drosophila melanogaster collected in three localities in Fukushima in 2012 and 2013 for chromosomal inversions. No unique inversion was found in 298 genomes in 2012 and only two in 2,006 genomes in 2013. The observed frequencies were even lower than the long-term average frequency of unique inversions in Japan. The common cosmopolitan inversions were also examined in Fukushima, Kyoto, and Iriomote (Okinawa) in 2012. Among three samples in Fukushima, the flies in Iizaka, where environmental radiation level was the highest, showed the lowest frequency of In(2L)t, but the highest frequency of In(3R)P, contrary to the expectation of decreasing of their frequencies in higher polluted areas. These results suggest that, at this level of genetic analysis, Fukushima populations of D. melanogaster would not have been negatively impacted following the release of radionuclides. Transposable P-element mobility was not likely to induce DNA damage solely or synergistically with radioactivity, because their transposition activity was totally repressed in the Fukushima strains. However, it should be noted that, because of limitations in access to the exclusion zone, we could only sample the populations in areas of relatively low radioactive contamination (0.39-0.63 μSv/h). Therefore, the present study is likely to be underpowered to detect any effects that might be expected in heavily contaminated areas.

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation

International Nuclear Information System (INIS)

Tentner, A.M.; Parma, E.; Wei, T.; Wigeland, R.

2010-01-01

An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

Severe accident approach - final report. Evaluation of design measures for severe accident prevention and consequence mitigation.

Energy Technology Data Exchange (ETDEWEB)

Tentner, A. M.; Parma, E.; Wei, T.; Wigeland, R.; Nuclear Engineering Division; SNL; INL

2010-03-01

An important goal of the US DOE reactor development program is to conceptualize advanced safety design features for a demonstration Sodium Fast Reactor (SFR). The treatment of severe accidents is one of the key safety issues in the design approach for advanced SFR systems. It is necessary to develop an in-depth understanding of the risk of severe accidents for the SFR so that appropriate risk management measures can be implemented early in the design process. This report presents the results of a review of the SFR features and phenomena that directly influence the sequence of events during a postulated severe accident. The report identifies the safety features used or proposed for various SFR designs in the US and worldwide for the prevention and/or mitigation of Core Disruptive Accidents (CDA). The report provides an overview of the current SFR safety approaches and the role of severe accidents. Mutual understanding of these design features and safety approaches is necessary for future collaborations between the US and its international partners as part of the GEN IV program. The report also reviews the basis for an integrated safety approach to severe accidents for the SFR that reflects the safety design knowledge gained in the US during the Advanced Liquid Metal Reactor (ALMR) and Integral Fast Reactor (IFR) programs. This approach relies on inherent reactor and plant safety performance characteristics to provide additional safety margins. The goal of this approach is to prevent development of severe accident conditions, even in the event of initiators with safety system failures previously recognized to lead directly to reactor damage.

Development of a prototype graphic simulation program for severe accident training

International Nuclear Information System (INIS)

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

2000-05-01

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

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.

ACCIDENT PHENOMENA OF RISK IMPORTANCE PROJECT - Continued RESEARCH CONCERNING SEVERE ACCIDENT PHENOMENA AND MANAGEMENT IN Sweden

International Nuclear Information System (INIS)

Rolandson, S.; Mueller, F.; Loevenhielm, G.

1997-01-01

Since 1988 all reactors in Sweden have mitigating measures, such as filtered vents, implemented. In parallel with the work of implementing these measures, a cooperation effort (RAMA projects) between the Swedish utilities and the Nuclear Power Inspectorate was performed to acquire sufficient knowledge about severe accident research work. The on-going project has the name Accident Phenomena of Risk Importance 3. In this paper, we will give background information about severe accident management in Sweden. In the Accident Phenomena of Risk Importance 3 project we will focus on the work concerning coolability of melted core in lower plenum which is the main focus of the In-vessel Coolability Task Group within the Accident Phenomena of Risk Importance 3 project. The Accident Phenomena of Risk Importance 3 project has joined on international consortium and the in-vessel cooling experiments are performed by Fauske and Associates, Inc. in Burr Ridge, Illinois, United States America, Sweden also intends to do one separate experiment with one instrument penetration we have in Swedish/Finnish BWR's. Other parts of the Accident Phenomena of Risk Importance 3 project, such as support to level 2 studies, the research at Royal Institute of Technology and participation in international programs, such as Cooperative Severe Accident Research Program, Advanced Containment Experiments and PHEBUS will be briefly described in the paper

Overview of severe accident research at KAERI

International Nuclear Information System (INIS)

Kim, H.D.; Kim, S.B.; Hong, S.W.; Kim, D.H.

2000-01-01

The severe accident research program at Korea Atomic Energy Research Institute, within the framework of governmental 10 year long-term nuclear R and D program, aims at the development of assessment techniques and accident management strategies for the prevention and mitigation of potential risk. The research program includes experimental efforts, development of phenomena specific models and development of an integrated computer code. The results of research program is intended to be utilized for the design of the advanced light water reactor and development of accident management strategies for the operating reactors. The main focused areas of recent investigation at KAERI are experiments on in-vessel core debris retention (SONATA-IV) and fuel coolant interaction (TROI) along with the development of models and integrated computer code (MIDAS). (author)

Dosimetric and biological assessment in the Cuban program for children from areas affected by the Chernobyl accidents

International Nuclear Information System (INIS)

Garcia Lima, O.; Cruz, R.; Valdez, M.; Cardenas, J.; Jova, L.; Lopez, G.; Arado, O.; Lamadrid, A.I.; Morera, L.

1996-01-01

Since 1990 Cuba has been developing health care program to offer highly specialized medical assistance and to develop a rehabilitation program for children affected by the Chernobyl accident. as part of the program a group of dosimetric and biomedical studies have been performed tom asses the impact of the accident. The results obtained from the study with 4500 children streets the fact that external radiation was the most important contributor to the total dose, which was estimated between 1-170 mSv during 70 years, for an equal period of time the 137C s internal dose ranged between 1.5 and 565 Bq/Kg. The biomedical studies have been carried out in groups formed according to the surface contamination of the field and to the 137C s activity measured, The thyroid hyperplasia tend to increase with the increment of both parameters. The other analyzed factors did not show differences among the groups

Radiation protection program for assistance of victims of radiation accidents

International Nuclear Information System (INIS)

Fajardo, P.W.; Costa Silva, L.H. da; Rosa, R.

1991-11-01

The principles aspects of a radiological protection program for hospitals in case of medical assistance to external and internal contaminated persons are showed. It is based on the experience obtained at Centro Medico Naval Marcilio Dias during the assistance to the victims of Goiania accident in 1987. This paper describes the basic infrastructure of a nursery and the radiation protection procedures for the access control of people and materials, area and personal monitoring, decontamination and the support activities such as calibration of radiation monitors and waste management. Is is also estimated the necessary radiation protection materials and the daily quantity of waste generated. (author)

Statistical evaluation of design-error related accidents

International Nuclear Information System (INIS)

Ott, K.O.; Marchaterre, J.F.

1980-01-01

In a recently published paper (Campbell and Ott, 1979), a general methodology was proposed for the statistical evaluation of design-error related accidents. The evaluation aims at an estimate of the combined residual frequency of yet unknown types of accidents lurking in a certain technological system. Here, the original methodology is extended, as to apply to a variety of systems that evolves during the development of large-scale technologies. A special categorization of incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of the nuclear power reactor technology, considering serious accidents that involve in the accident-progression a particular design inadequacy

ALWR severe accident issue resolution in support of updated emergency planning

International Nuclear Information System (INIS)

Additon, Stephen L.; Leaver, David E.; Sorrell, Steven W.; Theofanous, Theo G.

2004-01-01

. The severe accident risk characteristics of the ALWRs reflect an emphasis on accident prevention, which is quantified in the URD as a maximum permissible core damage frequency of less than one occurrence in 100,000 reactor years. For severe accident sequences of a frequency lower than this criterion, the URD safety policy requires provisions to arrest, mitigate, and contain the accident and, accordingly, opportunities to terminate a core melt sequence are provided whenever practical at every stage of core degradation. This includes design provisions to maximize the chances of success for reflooding the reactor by depressurizing the primary system, provisions to ensure retention of core debris in the reactor vessel by cooling the outside of the reactor vessel, and provisions for a more favorable geometry for core debris cooling in the reactor cavity in order to slow and then terminate a core-concrete interaction. For all risk-significant branches of the containment event tree, it must be demonstrated that early containment failure is avoided. This paper addresses the severe accident issue resolution tasks which were undertaken by the U.S. ALWR Program and ARSAP to ensure that the capability of passive ALWRs to arrest, mitigate and contain severe accidents would be sufficient to justify a significant change in the appropriate emergency planning requirements. The next section summarizes all of the issue resolution activities that will culminate in the issuance by the U.S. Nuclear Regulatory Commission (NRC) of a Final Safety Evaluation Report for the passive ALWR URD, scheduled for January 1994. The following section addresses more recent activities undertaken by ARSAP to enhance the issue resolution basis and to provide additional confirmatory evidence supporting the URD criteria. Included are the ongoing activities to establish a technical case, if possible, for in-vessel retention for the passive PWR and for the accommodation of ex-vessel steam explosions in the

Severe accident risks: An assessment for five US nuclear power plants

International Nuclear Information System (INIS)

1991-01-01

This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United State. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two of the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide releases and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. This report, Volume 3, contains two appendices. Appendix D summarizes comments received, and staff responses, on the first (February 1987) draft of NUREG-1150. Appendix E provides a similar summary of comments and responses, but for the second (June 1989) version of the report

Learning from nuclear accident experience

International Nuclear Information System (INIS)

Vaurio, J.K.

1984-01-01

Statistical procedures are developed to estimate accident occurrence rates from historical event records, to predict future rates and trends, and to estimate the accuracy of the rate estimates and predictions. Maximum likelihood estimation is applied to several learning models, and results are compared to earlier graphical and analytical estimates. The models are based on (1) the cumulative number of operating years, (2) the cumulative number of plants built, and (3) accidents (explicitly), with the accident rate distinctly different before and after an accident. The statistical accuracies of the parameters estimated are obtained in analytical form using the Fisher information matrix. Using data on core damage accidents in electricity producing plants, it is estimated that the probability for a plant to have a serious flaw has decreased from 0.1 to 0.01 during the developmental phase of the nuclear industry. At the same time the equivalent frequency of accidents has decreased from 0.04 per reactor year to 0.0004 per reactor year, partly due to the increasing population of plants. 10 references, 7 figures, 2 tables

Assessment of Mobile Accident Response Capability

International Nuclear Information System (INIS)

1983-03-01

This report presents the results of a DOE-sponsored assessment of nuclear accident response resources. It identifies the mobile resources that could be required to respond to different types of nuclear accidents including major ones like TMI-2, identifies the resources currently available and makes recommendations for the design and construction of additional mobile accident response resources to supplement those already in existence. This project is referred to as the Mobile Accident Response Capability (MARC) program

Research notes : are safety corridors really safe? Evaluation of the corridor safety improvement program.

Science.gov (United States)

1998-08-26

High accident frequencies on Oregons highway corridors are of concern to the Oregon Department of Transportation (ODOT). : ODOT adopted the Corridor Safety Improvement Program as part of an overall program of safety improvements using federal and ...

Development of a prototype graphic simulation program for severe accident training

Energy Technology Data Exchange (ETDEWEB)

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

2000-05-01

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

29 CFR 1960.70 - Reporting of serious accidents.

Science.gov (United States)

2010-07-01

... 29 Labor 9 2010-07-01 2010-07-01 false Reporting of serious accidents. 1960.70 Section 1960.70... accidents. Agencies must provide the Office of Federal Agency Programs with a summary report of each fatal and catastrophic accident investigation. The summaries shall address the date/time of accident, agency...

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

Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management.

Science.gov (United States)

Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

2016-04-01

Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8% of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced.

Analysis of Surface Water Pollution Accidents in China: Characteristics and Lessons for Risk Management

Science.gov (United States)

Yao, Hong; Zhang, Tongzhu; Liu, Bo; Lu, Feng; Fang, Shurong; You, Zhen

2016-04-01

Understanding historical accidents is important for accident prevention and risk mitigation; however, there are no public databases of pollution accidents in China, and no detailed information regarding such incidents is readily available. Thus, 653 representative cases of surface water pollution accidents in China were identified and described as a function of time, location, materials involved, origin, and causes. The severity and other features of the accidents, frequency and quantities of chemicals involved, frequency and number of people poisoned, frequency and number of people affected, frequency and time for which pollution lasted, and frequency and length of pollution zone were effectively used to value and estimate the accumulated probabilities. The probabilities of occurrences of various types based on origin and causes were also summarized based on these observations. The following conclusions can be drawn from these analyses: (1) There was a high proportion of accidents involving multi-district boundary regions and drinking water crises, indicating that more attention should be paid to environmental risk prevention and the mitigation of such incidents. (2) A high proportion of accidents originated from small-sized chemical plants, indicating that these types of enterprises should be considered during policy making. (3) The most common cause (49.8 % of the total) was intentional acts (illegal discharge); accordingly, efforts to increase environmental consciousness in China should be enhanced.

Investigation program on PWR-steel-containment behavior under accident conditions

International Nuclear Information System (INIS)

Krieg, R.; Eberle, F.; Goeller, B.; Gulden, W.; Kadlec, J.; Messemer, G.; Mueller, S.; Wolf, E.

1983-10-01

This report is a first documentation of the KfK/PNS activities and plans to investigate the behaviour of steel containments under accident conditions. The investigations will deal with a free standing spherical containment shell built for the latest type of a German pressurized water reactor. The diameter of the containment shell is 56 m. The minimum wall thickness is 38 mm. The material used is the ferritic steel 15MnNi63. According to the actual planning the program is concerned with four different problems which are beyond the common design and licensing practice: Containment behavior under quasi-static pressure increase up to containment failure. Containment behavior under high transient pressures. Containment oscillations due to earthquake loadings; consideration of shell imperfections. Containment buckling due to earthquake loadings. The investigation program consists of both theoretical and experimental activities including membrane tests allowing for very high plastic strains and oscillation tests with a thin-walled, high-accurate spherical shell. (orig.) [de

Genital endometriosis rate dynamics before and after Chernobyl accident

International Nuclear Information System (INIS)

Al-Shbul', I.; Suprun, L.Ya.

2000-01-01

The necessity of endometriosis dynamics evaluation is caused by worse ecological situation on the area of Belarus. Genital endometriosis frequency was studied considering the outcomes of surgeries fulfilled in hospitals of Gomel, Mogilev and Vitebsk in 1981-1995. At this time 1254 women underwent an operation and 19% of patients (235 persons) were operated before the Chernobyl accident. In the first 5 years after the accident endometriosis frequency increased nearly 2 times. The next 5 years (1991-1995) the number of operated patients was 565, i.e. 45% from the whole number. Uterus was extirpated or amputated in 898 patients, ovaries at both sides were removed in 36 ones. As the analysis showed the endometriosis frequency grew in 2,5 times for last 15 years, the most significant increase of this pathology was observed during the first five years after the accident

A structured approach to individual plant evaluation and accident management

International Nuclear Information System (INIS)

Klopp, G.T.

1991-01-01

The current requirements for the performance of individual plant evaluations (IPE's) include the derivation of accident management insights as and if they occur in the course of finalizing an IPE. The development of formal, structured accident management programs is, however, explicitly excluded from current IPE requirements. The Nuclear Regulatory Commission is following the Nuclear Management and Resources Council (NUMARC) efforts to establish the framework(s) for accident management program development and plants to issue requirements on such development at a later date. The Commonwealth Edison program consists of comprehensive level 2 PRA's which address the requirements for IPE's and which go beyond those requirements. From the start of the IPE efforts, it was firmly held, within Edison, that the best way to fully and economically extract a viable accident management program from an IPE was to integrate the two efforts from the start and include the accident management program development as a required IPE product

KADIS: a program to analyse the disassembly phase of hypothetical accidents in LMFBRs

International Nuclear Information System (INIS)

Schmuck, P.; Jacobs, G.; Arnecke, G.

1977-11-01

The program KADIS models the disassembly phase during power excursions in LMFBR hypothetical accidents. KADIS is based on point kinetics in the neutronics part and on a 2-dimensional representation of the reactor core in the hydrodynamics part. The core is modeled as an ideal, compressible fluid which is heated up adiabatically during the excursion. KADIS was built up with the help of the VENUS program of Argonne National Laboratory. Several important features were added to the basic VENUS model. Therefore we give first a complete description of the mathematical models used. Secondly we provide the user with the necessary information to handle the input/output of KADIS. (orig.) [de

Psychosomatic health status of children exposed to the Chernobyl accident

International Nuclear Information System (INIS)

Korol, N.; Shibata, Yoshisada; Nakane, Yoshibumi

1998-01-01

Childhood victims were investigated focussing on the psychosomatic disorders. The subjects were some of the 3834 children who evacuated from the Chernobyl zone to Kiev (evacuees) and 200 children who have been living in Kiev since prior to the accident (comparison group). A psychological test administered to 504 evacuees aged 12-14 years at the time of the accident and the comparison group indicated that the frequencies of neutroticism, high level of anxiety and conflicts were significantly higher in the evacuees than in the comparison group (p<0.001). Another psychological test administered at puberty to the 504 evacuees and 200 other evacuees exposed to the accident at 4-6 years of age indicated that the psycho-emotional portrait of evacuated teenagers significantly changed with time since the accident. The effects of the Chernobyl accident on the health of the vegetative dystonia observed in 1987-1990 and 1990-1995 were higher in the evacuees than in the comparison group, although they were not statistically significant. Furthermore, a significant (p<0.001) association of the vegetative dystonia with peptic and cardiovascular disorders was observed. The present study indicates that the vegetative dystonia is still highly prevalent among childhood victims and deems to support that the vegetative dystonia may be a precursor of several diseases such as cardiovascular and peptic disorders. It should be emphasized that a health promotion program to produce a change in psychological and social problems after the Chernobyl accident is necessary to decrease the health impact among Ukrainian people. (author)

Psychosomatic health status of children exposed to the Chernobyl accident

Energy Technology Data Exchange (ETDEWEB)

Korol, N. [Scientific Center for Radiation Medicine, Kiev (Ukraine); Shibata, Yoshisada; Nakane, Yoshibumi

1998-12-01

Childhood victims were investigated focussing on the psychosomatic disorders. The subjects were some of the 3834 children who evacuated from the Chernobyl zone to Kiev (evacuees) and 200 children who have been living in Kiev since prior to the accident (comparison group). A psychological test administered to 504 evacuees aged 12-14 years at the time of the accident and the comparison group indicated that the frequencies of neutroticism, high level of anxiety and conflicts were significantly higher in the evacuees than in the comparison group (p<0.001). Another psychological test administered at puberty to the 504 evacuees and 200 other evacuees exposed to the accident at 4-6 years of age indicated that the psycho-emotional portrait of evacuated teenagers significantly changed with time since the accident. The effects of the Chernobyl accident on the health of the vegetative dystonia observed in 1987-1990 and 1990-1995 were higher in the evacuees than in the comparison group, although they were not statistically significant. Furthermore, a significant (p<0.001) association of the vegetative dystonia with peptic and cardiovascular disorders was observed. The present study indicates that the vegetative dystonia is still highly prevalent among childhood victims and deems to support that the vegetative dystonia may be a precursor of several diseases such as cardiovascular and peptic disorders. It should be emphasized that a health promotion program to produce a change in psychological and social problems after the Chernobyl accident is necessary to decrease the health impact among Ukrainian people. (author)

Predicted occurrence rate of severe transportation accidents involving large casks

International Nuclear Information System (INIS)

Dennis, A.W.

1978-01-01

A summary of the results of an investigation of the severities of highway and railroad accidents as they relate to the shipment of large radioactive materials casks is discussed. The accident environments considered are fire, impact, crash, immersion, and puncture. For each of these environments, the accident severities and their predicted frequencies of occurrence are presented. These accident environments are presented in tabular and graphic form to allow the reader to evaluate the probabilities of occurrence of the accident parameter severities he selects

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

International Nuclear Information System (INIS)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M.; Bley, D.; Johnson, D.

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Analysis of core damage frequency from internal events during mid-loop operations. Appendix E (Sections E.9-E.16), Volume 2, Part 3B

Energy Technology Data Exchange (ETDEWEB)

Chu, T.L.; Musicki, Z.; Kohut, P.; Yang, J.; Bozoki, G.; Hsu, C.J.; Diamond, D.J.; Wong, S.M. [Brookhaven National Lab., Upton, NY (United States); Bley, D.; Johnson, D. [PLG Inc., Newport Beach, CA (United States)] [and others

1994-06-01

Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analyses that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The scope of the level-1 study includes plant damage state analysis, and uncertainty analysis. Volume 1 summarizes the results of the study. Internal events analysis is documented in Volume 2. It also contains an appendix that documents the part of the phase 1 study that has to do with POSs other than mid-loop operation. Internal fire and internal flood analyses are documented in Volumes 3 and 4. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associates, Inc. Volume 6 documents the accident progression, source terms, and consequence analysis.

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

Compilation of accident statistics in PSE

International Nuclear Information System (INIS)

Jobst, C.

1983-04-01

The objective of the investigations on transportation carried out within the framework of the 'Project - Studies on Safety in Waste Management (PSE II)' is the determination of the risk of accidents in the transportation of radioactive materials by rail. The fault tree analysis is used for the determination of risks in the transportation system. This method offers a possibility for the determination of frequency and consequences of accidents which could lead to an unintended release of radionuclides. The study presented compiles all data obtained from the accident statistics of the Federal German Railways. (orig./RB) [de

Revised GCFR safety program plan

International Nuclear Information System (INIS)

Kelley, A.P.; Boyack, B.E.; Torri, A.

1980-05-01

This paper presents a summary of the recently revised gas-cooled fast breeder reactor (GCFR) safety program plan. The activities under this plan are organized to support six lines of protection (LOPs) for protection of the public from postulated GCFR accidents. Each LOP provides an independent, sequential, quantifiable risk barrier between the public and the radiological hazards associated with postulated GCFR accidents. To implement a quantitative risk-based approach in identifying the important technology requirements for each LOP, frequency and consequence-limiting goals are allocated to each. To ensure that all necessary tasks are covered to achieve these goals, the program plan is broken into a work breakdown structure (WBS). Finally, the means by which the plan is being implemented are discussed

Severe accident sequence assessment for boiling water reactors: program overview

International Nuclear Information System (INIS)

Fontana, M.H.

1980-10-01

The Severe Accident Sequence Assessment (SASA) Program was started at the Oak Ridge National Laboratory (ORNL) in June 1980. This report documents the initial planning, specification of objectives, potential uses of the results, plan of attack, and preliminary results. ORNL was assigned the Brown's Ferry Unit 1 Plant with the station blackout being the initial sequence set to be addressed. This set includes: (1) loss of offsite and onsite ac power with no coolant injection; and (2) loss of offsite and onsite ac power with high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC) as long as dc power supply lasts. This report includes representative preliminary results for the former case

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.)

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

International Nuclear Information System (INIS)

Hu Erbang; Gao Zhanrong; Zhang Heyuan; Wei Weiqiang

1996-12-01

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

Differences in frequency rates and classes of occupational accidents in copper ore and black coal mines

Energy Technology Data Exchange (ETDEWEB)

Stecko, R.; Zacharzewski, J.

1987-02-01

This paper analyzes occupational accidents in copper mines and black coal mines in Poland from 1968 to 1985. Evaluates accident rate calculated per 1000 employees and per 100,000 t coal or copper ore output and distribution of accident classes characterized on the basis of miner absenteeism caused by an accident. Accident rate in coal mines calculated per 1,000 miners was: 58.5 in 1981, 51.0 in 1982, 39.5 in 1983, 36.6 in 1984 and 32.1 in 1985. Accident rate calculated per 100,000 t coal output was: 5.8 in 1980, 13.7 in 1981, 10.7 in 1982, 8.2 in 1983, 7.8 in 1984 and 6.7 in 1985. Curves showing accident rates and indices characterizing accident classes in 1968-1985 are plotted. Comparative evaluations of accident rate fluctuations in copper and coal mines suggest that these fluctuations are associated with causes other than mining and geologic conditions. Certain similiarities in accident rate fluctuations could be determined in spite of differing mining and geologic conditions and mine service life in the Lubin copper basin and Upper Silesia coal basin. 4 refs.

40 CFR 68.168 - Five-year accident history.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Five-year accident history. 68.168 Section 68.168 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY (CONTINUED) AIR PROGRAMS (CONTINUED) CHEMICAL ACCIDENT PREVENTION PROVISIONS Risk Management Plan § 68.168 Five-year accident history...

Accident management information needs

International Nuclear Information System (INIS)

Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R.

1990-04-01

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

Accident management information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Ward, L.W.; Nelson, W.R.; Meyer, O.R. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1990-04-01

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

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.

Proposal for computer investigation of LMFBR core meltdown accidents

International Nuclear Information System (INIS)

Boudreau, J.E.; Harlow, F.H.; Reed, W.H.; Barnes, J.F.

1974-01-01

The environmental consequences of an LMFBR accident involving breach of containment are so severe that such accidents must not be allowed to happen. Present methods for analyzing hypothetical core disruptive accidents like a loss of flow with failure to scram cannot show conclusively that such accidents do not lead to a rupture of the pressure vessel. A major deficiency of present methods is their inability to follow large motions of a molten LMFBR core. Such motions may lead to a secondary supercritical configuration with a subsequent energy release that is sufficient to rupture the pressure vessel. The Los Alamos Scientific Laboratory proposes to develop a computer program for describing the dynamics of hypothetical accidents. This computer program will utilize implicit Eulerian fluid dynamics methods coupled with a time-dependent transport theory description of the neutronic behavior. This program will be capable of following core motions until a stable coolable configuration is reached. Survey calculations of reactor accidents with a variety of initiating events will be performed for reactors under current design to assess the safety of such reactors

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).

Overview of severe accident research at the USNRC

International Nuclear Information System (INIS)

Basu, S.; Ader, C.E.

1999-01-01

This paper summarizes the U.S. Nuclear Regulatory Commission's (USNRC) severe accident research activities, in particular, progress made in the past year toward the resolution and/or improved understanding of a number of severe accident issues. The direct containment heating (DCH) is nearing resolution for Combustion Engineering and Babcock and Wilcox type pressurized water reactors (PWRs) are well as for ice condensers. Additionally, two lower pressure DCH tests were conducted recently at the Sandia National Laboratories (SNL) under the NRC/IPSN/FzK sponsorship to provide data regarding intentional depressurization as an accident management strategy to mitigate DCH loads. In the area of lower head integrity, the experimental program to investigate boiling heat transfer on downward facing curved surfaces with insulation was completed. Finally, the SNL program investigating the creep rupture behavior of the lower head under the combined thermo-mechanical loading was completed recently. Additional lower head experiments at SNL are being planned as an OECD project. During the past year, the USNRC participated in two programs aimed at extending the data base on hydrogen combustion into more prototypic situations. Testing was performed at the Brookhaven National Laboratory (BNL) to investigate detonation transmission at elevated temperatures. In a cooperative program under the sponsorship of NRC/IPSN/FzK, Russian Research Center (RRC) investigated hydrogen combustion issues at large scale at the RUT facility. The experimental program at the SNL to examine the performance of Passive Autocatalytic Recombiners (PARs) was completed also this year. In the fuel-coolant interaction (FCI) area, the experimental work at the Argonne National Laboratory (ANL) to investigate chemical augmentation of FCI energetics was completed as was the experimental work at the University of Wisconsin (UW) involving one-dimensional propagation experiments (similar to KROTOS). The USNRC is

Accident prevention in power plants

International Nuclear Information System (INIS)

Steyrer, H.

Large thermal power plants are insured to a great extent at the Industrial Injuries Insurance Institute of Instrument and Electric Engineering. Approximately 4800 employees are registered. The accident frequency according to an evaluation over 12 months lies around 79.8 per year and 1000 employees in fossil-fired power plants, around 34.1 per year and 1000 employees in nuclear power plants, as in nuclear power plants coal handling and ash removal are excluded. Injuries due to radiation were not registered. The crucial points of accidents are mechanical injuries received on solid, sharp-edged and pointed objects (fossil-fired power plants 28.6%, nuclear power plants 41.5%), stumbling, twisting or slipping (fossil-fired power plants 21.8%, nuclear power plants 19.5%) and injuries due to moving machine parts (only nuclear power plants 12.2%). However, accidents due to burns or scalds obtain with 4.2% and less a lower portion than expected. The accident statistics can explain this fact in a way that the typical power plant accident does not exist. (orig./GL) [de

Zagreb and Tenerife: Airline Accidents Involving Linguistic Factors

Science.gov (United States)

Cookson, Simon

2009-01-01

The International Civil Aviation Organization (ICAO) is currently implementing a program to improve the language proficiency of pilots and air traffic controllers worldwide. In justifying the program, ICAO has cited a number of airline accidents that were at least partly caused by language factors. Two accidents cited by ICAO are analysed in this…

The Epidemiology of Injuries and Accidents in Children Under one Year of Age, during (2009-2016 in Hamadan Province, Iran

Directory of Open Access Journals (Sweden)

Zaher Khazaei

2016-07-01

Full Text Available Background Injuries and accidents are the first cause of death in the first 5 years of children life in the world; the present study was conducted to investigate the extent and distribution of accidents in the infants under one year in Hamadan Province, Iran. Materials and Methods:This cross-sectional descriptive-analytic study was carried out using of   data of injuries and accidents related to children under one year for Hamadan province in seven years period from March 2009 to March 2016. In this study we used data according national injuries and accidents recorded program. The data were analyzed using descriptive statistics as well as analytical statistics including the Chi-square test. Data were analyzed using Stata software version 12. Results:In this 7 year periods, 3,200 accidents were registered among children under one year. The highest occurrence of accidents was in the spring 1,029 (31.15% of cases. 1,890 (59.1% of accidents occurred in the urban area and only 429 (13.4% of them were in rural area. In total, car accidents (53.4%, trauma (12.6% and fall from altitude (8.8% had most frequency from all accidents. There was a significant difference between gender and place of accident with type of accident (P

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

1991-01-01

The use of NUREG-1150 and similar probabilistic risk assessments in the Nuclear Regulatory Commission (NRC) and industry risk management programs is discussed. Risk management is more comprehensive than the commonly used term accident management. Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.; Sype, T.; Camp, A.; Maloney, K.

1990-01-01

The use of NUREG-1150 and similar Probabilistic Risk Assessments in NRC and industry risk management programs is discussed. ''Risk management'' is more comprehensive than the commonly used term ''accident management.'' Accident management includes strategies to prevent vessel breach, mitigate radionuclide releases from the reactor coolant system, and mitigate radionuclide releases to the environment. Risk management also addresses prevention of accident initiators, prevention of core damage, and implementation of effective emergency response procedures. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. Examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management are discussed, with particular attention given to the early phases of accidents. Plans and methods for evaluating accident management strategies that have been identified in the NRC accident management program are discussed. 2 refs., 3 figs

The public whole body counting program following the Three Mile Island accident. Technical report, April-September 1979

International Nuclear Information System (INIS)

Gotchy, R.L.; Bores, R.J.

1980-12-01

In early April, 1979 the U.S. Nuclear Regulatory instituted a program to determine whether any radioactivity released as a result of the March 28, 1979 accident at the Three Mile Island Unit-2 was accumulating in members of the general public living near Unit-2. The program used a device called a whole body counter which has the capability of measuring very small quantities of radioactivity in people. There were 753 men, women and children successfully counted; nine of these people were counted a second time, leading to a total of 762 whole body counts. There was no radioactivity identified in any member of the public which could have originated from the radioactive materials released following the accident. Several people with higher than average levels of naturally occurring radioactivity were identified. The counting systems used are briefly described. Technical problems encountered, results and conclusions are discussed

NPP Krsko Severe Accident Management Guidelines Implementation

International Nuclear Information System (INIS)

Basic, I.; Krajnc, B.; Bilic-Zabric, T.; Spiler, J.

2002-01-01

Severe Accident Management is a framework to identify and implement the Emergency Response Capabilities that can be used to prevent or mitigate severe accidents and their consequences. The USA NRC has indicated that the development of a licensee plant specific accident management program will be required in order to close out the severe accident regulatory issue (Ref. SECY-88-147). Generic Letter 88-20 ties the Accident management Program to IPE for each plant. The SECY-89-012 defines those actions taken during the course of an accident by the plant operating and technical staff to: 1) prevent core damage, 2) terminate the progress of core damage if it begins and retain the core within the reactor vessel, 3) maintain containment integrity as long as possible, and 4) minimize offsite releases. The subject of this paper is to document the severe accident management activities, which resulted in a plant specific Severe Accident Management Guidelines implementation. They have been developed based on the Krsko IPE (Individual Plant Examination) insights, Generic WOG SAMGs (Westinghouse Owners Group Severe Accident Management Guidances) and plant specific documents developed within this effort. Among the required plant specific actions the following are the most important ones: Identification and documentation of those Krsko plant specific severe accident management features (which also resulted from the IPE investigations). The development of the Krsko plant specific background documents (Severe Accident Plant Specific Strategies and SAMG Setpoint Calculation). Also, paper discusses effort done in the areas of NPP Krsko SAMG review (internal and external ), validation on Krsko Full Scope Simulator (Severe Accident sequences are simulated by MAAP4 in real time) and world 1st IAEA Review of Accident Management Programmes (RAMP). (author)

Monitoring of congenital malformations in Belarus after the Chernobyl accident

International Nuclear Information System (INIS)

Lazjuk, G.I.; Kirillova, I.A.; Nikolaev, D.L.; Novikova, I.V.

1993-01-01

An investigation of over 21,000 embryos and fetuses from medically-induced abortions was conducted from 1980 through 1991 in the Republic of Belarus. More than half of the abortions studied were carried out after the Chernobyl nuclear accident, including 1176 from districts with 137 Cs soil contamination levels over 0.6 TBq/km 2 (15 Ci/km 2 ). Congenital malformations (CM's) in 7325 newborn children also were analyzed. The data on these children were obtained from a genetic monitoring program. It was shown that in the 5 years after the Chernobyl accident the frequency of abnormal developments in aborted fetuses from contaminated areas was significantly higher than in aborted fetuses from Minsk, which was relatively uncontaminated. Additionally, the CM incidence in newborn children increased in Belarus compared to the CM incidences before the accident; the increase was most significant in the heavily contaminated areas. The increases were attributed primarily to CMS characterized by dominant mutations. These increases could have been partially caused by factors unrelated to radiation dose, including defective nourishment, chemical contaminants, and psychological stresses. A correlation between CM increase and the parents' dose has not been established. 17 refs., 6 tabs

Accident management strategy in Sweden - implementation and verification

International Nuclear Information System (INIS)

Loewenhielm, Gustaf; Engqvist, Alf; Espefaelt, Ralf

1994-01-01

A comprehensive program for severe accident mitigation was completed in Sweden by the end of 1988. As described in this paper, this program included plant modifications such as the introduction of filtered containment venting, and an accident management system comprising emergency operating strategies and procedures, training and emergency drills. The accident management system at Vattenfall has been further developed since 1988 and some results and experience from this development are reported in this paper. The main aspects covered concern the emergency organization and the supporting tools developed for use by the emergency response teams, the radiological implications such as accessibility to various locations and the long-term aspects of accident management. ((orig.))

An experience of information support of the Russian federal programs of the overcoming of consequences of the accident at the Chernobyl NPP

International Nuclear Information System (INIS)

Linge, I.I.; Ossipiants, I.A.; Ilushkin, I.I.; Melihova, E.I.; Blinov, B.K.; Marchenko, T.A.

2002-01-01

Since 1992, by a number of paragraphs of the federal programs on overcoming consequences of accident on Chernobyl NPP, the measures on informational-analytic support of the federal programs were provided. Within the framework of this activity for the solution of various aspects of Chernobyl's problem central bank of generalized data and numerous information systems were created. In the report the brief description of some of them is presented. In particular the databank on radioactive situation includes the information on 12 thousand settlements of Russia which have been exposed to radioactive contamination as a result of the accident.The medico-demographic section of a databank includes information on death rate for the reasons for all subjects of Russian Federation since 1982 till the present time. The developed information systems are available to all participants of work on overcoming of consequences of the accident. There are given the examples of integral estimates in short- and long-term forecasts of development of a situation in territories suffered by the accident at the Chernobyl NPP. (author)

Development of a new Frequency Program in the CERN Proton Synchrotron

CERN Document Server

Sundal, Magnus; Larsen, Ragnhild; Johnsen, Ragnar

The development of a new frequency program in the Proton Synchrotron (PS) is a project that links existing and new infrastructure in the first particle accelerator built at CERN. In a particle synchrotron, the magnetic field strength of the bending magnets and the revolution frequency of the particles are coupled to keep a circulating beam at fixed orbit during acceleration. The frequency program delivers a revolution frequency based on the measurements of the bending magnets. Ongoing renovations in the magnet measurement system are changing the way the information is distributed from the magnets. Once implemented, it will represent a step towards fully digital beam controls. This project involves development of both Very High Speed Integrated Circuit Hardware Description Language (VHDL) firmware and Printed Circuit Boards (PCBs) in the form of FPGA Mezzanine Cards (FMCs) to adapt the frequency program to the new transmission protocol called White Rabbit. The White Rabbit is a network solution that provides s...

Design and Development of a Severe Accident Training System

International Nuclear Information System (INIS)

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

2005-01-01

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

Discussion on several issues of the accidents management of nuclear power plants in operation

International Nuclear Information System (INIS)

Cao Xuewu; Wang Zhe; Zhang Yingzhen

2009-01-01

This article discusses several issues of the accident management of nuclear power plants in operation, for example: the necessity, implementation principle of accident management and accident management program etc. For conducting accident management for beyond design basis accidents, this article thinks that the accident management program should be developed and implemented to ensure that the plant and its personnel with responsibilities for accident management are adequately prepared to take effective on-site actions to prevent or mitigate the consequences of severe accident. (authors)

Leakage potential through mechanical penetrations in a severe accident environment

International Nuclear Information System (INIS)

Koenig, L.N.

1986-01-01

This paper reviews the findings of an ongoing program, Integrity of Containment Penetrations Under Severe Accident Loads. The program is concerned with the leakage modes as well as the magnitude of leakage through mechanical penetrations in a containment building subject to a severe accident. Seal and gasket tests are used to evaluate the effect of radiation aging, thermal aging, seal geometry, and seal squeeze on seals and gaskets subjected to a hypothesized severe accident. The effects on leakage of the structural response of equipment hatches, personnel airlocks, and drywell heads subjected to severe accident pressures are studied by experiments and analyses. The data gathered during this program will be used to develop methodologies for predicting leakage

Statistical evaluation of design-error related nuclear reactor accidents

International Nuclear Information System (INIS)

Ott, K.O.; Marchaterre, J.F.

1981-01-01

In this paper, general methodology for the statistical evaluation of design-error related accidents is proposed that can be applied to a variety of systems that evolves during the development of large-scale technologies. The evaluation aims at an estimate of the combined ''residual'' frequency of yet unknown types of accidents ''lurking'' in a certain technological system. A special categorization in incidents and accidents is introduced to define the events that should be jointly analyzed. The resulting formalism is applied to the development of U.S. nuclear power reactor technology, considering serious accidents (category 2 events) that involved, in the accident progression, a particular design inadequacy. 9 refs

Use of NUREG-1150 and IPEs in accident management

International Nuclear Information System (INIS)

Mauersberger

1992-01-01

The fundamental objective of the accident management program is to assure, in the event of a severe accident at a nuclear plant, that the effectiveness of personnel and equipment is maximized in preventing or mitigating the consequences of the accident. This document studies the use of NUREG-1150 and IPEs in accident management. Figs

Identification of NPP accidents using support vector classification

Energy Technology Data Exchange (ETDEWEB)

Back, Ju Hyun; Yoo, Kwae Hwan; Na, Man Gyun [Chosun University, Gwangju (Korea, Republic of)

2016-10-15

In case of the accidents that happens in a nuclear power plants (NPPs), it is very important to identify its accidents for the operator. Therefore, in order to effectively manage the accidents, the initial short time trends of major parameters have to be observed and NPP accidents have to accurately be identified to provide its information to operators and technicians. In this regard, the objective of this study is to identify the accidents when the accidents happen in NPPs. In this study, we applied the support vector classification (SVC) model to classify the initiating events of critical accidents such as loss of coolant accidents (LOCA), total loss of feedwater (TLOFW), station blackout (SBO), and steam generator tube rupture (SGTR). Input variables were used as the initial integral value of the signal measured in the reactor coolant system (RCS), steam generator, and containment vessel after reactor trip. The proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. In this study, the proposed SVC model is verified by using the simulation data of the modular accident analysis program (MAAP4) code. We used an initial integral value of the simulated sensor signals to identify the NPP accidents. The training data was used to train the SVC model. And, the trained model was confirmed using the test data. As a result, it was known that it can accurately classify five events.

RESULTS OF IMPLEMENTATION OF TARGET PROGRAMS ON RADIATION ACCIDENT REMEDIATION FOR THE PERIOD TO 2010 AND PROSPECTS OF FURTHER ACTIVITIES

Directory of Open Access Journals (Sweden)

T. A. Marchenko

2008-01-01

Full Text Available The report contains information about measures undertaken by the Ministry of Civil Defense and Emergencies of the Russian Federation in the framework of implementation of the state policy in the field of radiation accidents remediation. Results of works realized in the framework of target programs on remediation of radiation accidents at Chernobyl NPP and Production Association MAYAK, and on problems caused by nuclear weapon tests at Seminalatinsk test site are presented.

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-01-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

Uranium storage bed accident hazards evaluation

International Nuclear Information System (INIS)

Longhurst, G.R.; Shmayda, W.T.

1989-10-01

To properly assess hazards and risks associated with the use of uranium beds as tritium storage devices in fusion reactor systems, it is necessary to understand the consequences occurring in the event of an accident. Accidents involving uranium beds are postulated, and the possible results are considered. A research program to more fully and accurately understand those results has been initiated involving the Idaho National Engineering Laboratory and Ontario Hydro. The plan and objectives of that program are presented. 11 refs., 1 tab

[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)

The program system UFOMOD for assessing the consequences of nuclear accidents

International Nuclear Information System (INIS)

Ehrhardt, J.; Burkart, K.; Hasemann, I.; Matzerath, C.; Panitz, H.J.; Steinhauer, C.

1988-10-01

The programm system UFOMOD is a completely new accident consequence assessment (ACA) code. Its structure and modelling is based on the experience gained from applications of the old UFOMOD code during and after the German Risk Study - Phase A, the results of scientific investigations performed within the ongoing Phase B and the CEC-project MARIA, and the requirements resulting from the extended use of ACAs to help in decision-making. One of the most important improvements is the introduction of different trajecotry models for describing atmospheric dispersion in the near range and at larger distances. Emergency actions and countermeasures modelling takes into account recommendations of international commissions. The dosimetric models contain completely new age-, sex- and time-dependent data of dose-conversion factors for external and internal radiation; the ingestion pathway is modelled to consider seasonal dependencies. New dose-risk-relationships for stochastic and non-stochastic health effects are implemented; a special algorithm developed for ACA codes allows individual and collective leukemia and cancer risks to be presented as a function of time after the accident. According to the modular structure of the new program system UFOMOD, an easy access to parameter values and the results of the various submodels exists what facilitates sensitivity and uncertainty analyses. (orig.) [de

Uncertainties and severe-accident management

International Nuclear Information System (INIS)

Kastenberg, W.E.

1991-01-01

Severe-accident management can be defined as the use of existing and or alternative resources, systems, and actions to prevent or mitigate a core-melt accident. Together with risk management (e.g., changes in plant operation and/or addition of equipment) and emergency planning (off-site actions), accident management provides an extension of the defense-indepth safety philosophy for severe accidents. A significant number of probabilistic safety assessments have been completed, which yield the principal plant vulnerabilities, and can be categorized as (a) dominant sequences with respect to core-melt frequency, (b) dominant sequences with respect to various risk measures, (c) dominant threats that challenge safety functions, and (d) dominant threats with respect to failure of safety systems. Severe-accident management strategies can be generically classified as (a) use of alternative resources, (b) use of alternative equipment, and (c) use of alternative actions. For each sequence/threat and each combination of strategy, there may be several options available to the operator. Each strategy/option involves phenomenological and operational considerations regarding uncertainty. These include (a) uncertainty in key phenomena, (b) uncertainty in operator behavior, (c) uncertainty in system availability and behavior, and (d) uncertainty in information availability (i.e., instrumentation). This paper focuses on phenomenological uncertainties associated with severe-accident management strategies

Regulatory approach to enhanced human performance during accidents

International Nuclear Information System (INIS)

Palla, R.L. Jr.

1990-01-01

It has become increasingly clear in recent years that the risk associated with nuclear power is driven by human performance. Although human errors have contributed heavily to the two core-melt events that have occurred at power reactors, effective performance during an event can also prevent a degraded situation from progressing to a more serious accident, as in the loss-of-feedwater event at Davis-Besse. Sensitivity studies in which human error rates for various categories of errors in a probabilistic risk assessment (PRA) were varied confirm the importance of human performance. Moreover, these studies suggest that actions taken during an accident are at least as important as errors that occur prior to an initiating event. A program that will lead to enhanced accident management capabilities in the nuclear industry is being developed by the US Nuclear Regulatory Commission (NRC) and industry and is a key element in NRC's integration plan for closure of severe-accident issues. The focus of the accident management (AM) program is on human performance during accidents, with emphasis on in-plant response. The AM program extends the defense-in-depth principle to plant operating staff. The goal is to take advantage of existing plant equipment and operator skills and creativity to find ways to terminate accidents that are beyond the design basis. The purpose of this paper is to describe the NRC's objectives and approach in AM as well as to discuss several human performance issues that are central to AM

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.

Accident risks in the energy sector

International Nuclear Information System (INIS)

Burgherr, P.

2005-01-01

This article discusses the accident rate of natural gas installations, which are quoted by the author to be lowest of all fossil fuels. The statistics on accidents and their consequences are looked at for the whole natural gas supply chain. The results of a study commissioned by the Swiss Gas and Water Professionals Association (SVGW) are presented and discussed. Statistics for the European Union and Eastern Europe are looked at and analysed. The study's methodological basis is described and the criteria used for the definition of an accident considered to be 'serious' are listed. The results of comparisons made of various energy chains are presented and discussed. Graphics are presented of frequency of occurrence and seriousness of damage for various forms of energy as well as for maximum possible consequences of accidents. Specific analyses for the natural gas chain are presented

Interim reliability evaluation program (IREP)

International Nuclear Information System (INIS)

Carlson, D.D.; Murphy, J.A.

1981-01-01

The Interim Reliability Evaluation Program (IREP), sponsored by the Office of Nuclear Regulatory Research of the US Nuclear Regulatory Commission, is currently applying probabilistic risk analysis techniques to two PWR and two BWR type power plants. Emphasis was placed on the systems analysis portion of the risk assessment, as opposed to accident phenomenology or consequence analysis, since the identification of risk significant plant features was of primary interest. Traditional event tree/fault tree modeling was used for the analysis. However, the study involved a more thorough investigation of transient initiators and of support system faults than studies in the past and substantially improved techniques were used to quantify accident sequence frequencies. This study also attempted to quantify the potential for operator recovery actions in the course of each significant accident

A preliminary study for the implementation of general accident management strategies

International Nuclear Information System (INIS)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung

1997-01-01

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated

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

International Nuclear Information System (INIS)

Felix, E.; Dessars, N.

2003-01-01

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

Occupational Accidents with Agricultural Machinery in Austria.

Science.gov (United States)

Kogler, Robert; Quendler, Elisabeth; Boxberger, Josef

2016-01-01

The number of recognized accidents with fatalities during agricultural and forestry work, despite better technology and coordinated prevention and trainings, is still very high in Austria. The accident scenarios in which people are injured are very different on farms. The common causes of accidents in agriculture and forestry are the loss of control of machine, means of transport or handling equipment, hand-held tool, and object or animal, followed by slipping, stumbling and falling, breakage, bursting, splitting, slipping, fall, and collapse of material agent. In the literature, a number of studies of general (machine- and animal-related accidents) and specific (machine-related accidents) agricultural and forestry accident situations can be found that refer to different databases. From the database Data of the Austrian Workers Compensation Board (AUVA) about occupational accidents with different agricultural machinery over the period 2008-2010 in Austria, main characteristics of the accident, the victim, and the employer as well as variables on causes and circumstances by frequency and contexts of parameters were statistically analyzed by employing the chi-square test and odds ratio. The aim of the study was to determine the information content and quality of the European Statistics on Accidents at Work (ESAW) variables to evaluate safety gaps and risks as well as the accidental man-machine interaction.

Preprocessor for RELAP5 code, nuclear reactor thermal hydraulics accident analysis program, using Microsoft MS-EXCEL tool

International Nuclear Information System (INIS)

Biaty, Patricia Andrea Paladino; Sabundjian, Gaiane

2005-01-01

The thermal hydraulic study in accidents and transients analyses in nuclear power plants is realized with some special tools. These programs use the best estimate analyses and have been developed to simulate accidents and transients in Pressurized Water Reactors (PWR) and auxiliary systems. The RELAP5 code has been used as tool to licensing the nuclear facilities in our country, which is the objective of this study. The main problem when RELAP5 code is used is a lot of information necessary to simulate thermal hydraulic accidents. Moreover, there is the necessity of a reasonable amount of mathematical operations to calculation of the geometry of the components existents. Therefore, in order to facilitate the manipulation of this information, it is necessary the developing a friendly preprocessor for attainment of the mathematical calculations for RELAP5 code. One of the tools used for some of these calculations is the MS-EXCEL, which will be used in this work. (author)

FN-curves: preliminary estimation of severe accident risks after Fukushima

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Costa, Antonio Carlos Lopes da

2015-01-01

Doubts of whether the risks related to severe accidents in nuclear reactors are indeed very low were raised after the nuclear accident at Fukushima Daiichi in 2011. Risk estimations of severe accidents in nuclear power plants involve both probability and consequence assessment of such events. Among the ways to display risks, risk curves are tools that express the frequency of exceeding a certain magnitude of consequence. Societal risk is often represented graphically in a FN-curve, a type of risk curve, which displays the probability of having N or more fatalities per year, as a function of N, on a double logarithmic scale. The FN-curve, originally introduced for the assessment of the risks in the nuclear industry through the U.S.NRC Reactor Safety Study WASH-1400 (1975), is used in various countries to express and limit risks of hazardous activities. This first study estimated an expected rate of core damage equal to 5x10 -5 by reactor-year and suggested an upper bound of 3x10 -4 by reactor-year. A more recent report issued by Electric Power Research Institute - EPRI (2008) estimates a figure of the order of 2x10 -5 by reactor-year. The Fukushima nuclear accident apparently implies that the observed core damage frequency is higher than that predicted by these probabilistic safety assessments. Therefore, this paper presents a preliminary analyses of the FN-curves related to severe nuclear reactor accidents, taking into account a combination of available data of past accidents, probability modelling to estimate frequencies, and expert judgments. (author)

FN-curves: preliminary estimation of severe accident risks after Fukushima

Energy Technology Data Exchange (ETDEWEB)

Vasconcelos, Vanderley de; Soares, Wellington Antonio; Costa, Antonio Carlos Lopes da, E-mail: vasconv@cdtn.br, E-mail: soaresw@cdtn.br, E-mail: aclc@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEN-MG), Belo Horizonte, MG (Brazil)

2015-07-01

Doubts of whether the risks related to severe accidents in nuclear reactors are indeed very low were raised after the nuclear accident at Fukushima Daiichi in 2011. Risk estimations of severe accidents in nuclear power plants involve both probability and consequence assessment of such events. Among the ways to display risks, risk curves are tools that express the frequency of exceeding a certain magnitude of consequence. Societal risk is often represented graphically in a FN-curve, a type of risk curve, which displays the probability of having N or more fatalities per year, as a function of N, on a double logarithmic scale. The FN-curve, originally introduced for the assessment of the risks in the nuclear industry through the U.S.NRC Reactor Safety Study WASH-1400 (1975), is used in various countries to express and limit risks of hazardous activities. This first study estimated an expected rate of core damage equal to 5x10{sup -5} by reactor-year and suggested an upper bound of 3x10{sup -4} by reactor-year. A more recent report issued by Electric Power Research Institute - EPRI (2008) estimates a figure of the order of 2x10{sup -5} by reactor-year. The Fukushima nuclear accident apparently implies that the observed core damage frequency is higher than that predicted by these probabilistic safety assessments. Therefore, this paper presents a preliminary analyses of the FN-curves related to severe nuclear reactor accidents, taking into account a combination of available data of past accidents, probability modelling to estimate frequencies, and expert judgments. (author)

Road accidents involving bicycles: configurations and injuries.

Science.gov (United States)

Orsi, Chiara; Montomoli, Cristina; Otte, Dietmar; Morandi, Anna

2017-12-01

This study analyzed the most common types of accident involving bicycles and compared the frequency of injuries. The data source was the database of German In-Depth Accident Study (GIDAS). Cases consist of bicycles and their riders involved in accidents between 2000 and 2010. In most collisions, the bicycle impacted with a car. The percentage of injured bicyclists was higher in collisions with a heavy vehicle and decreased when the bicycle impacted with lighter vehicles. A high percentage of injured bicyclists in single accidents was observed; the most severe injury was more frequently to head and extremities. Accidents involving a car and a bicycle with the right of way in a bicycle path represented about 20% of involved and injured bicyclists. The ten most frequent configurations represented about 60% of involved and injured bicyclists. These results contribute to understand the dangerous scenarios for bicyclists and to suggest preventive actions.

Neural network-based expert system for severe accident management

International Nuclear Information System (INIS)

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

1992-01-01

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

Current state of the communications program concerning nuclear energy following the JCO accident

International Nuclear Information System (INIS)

Mitsui, Rie

2001-01-01

Full text: Public trust in nuclear energy was significantly affected by a criticality accident that occurred in September 1999 at JCO's uranium re-conversion plant in Tokai-mura Village, Ibaraki Prefecture, some 120 kilometres north-east of Tokyo. An opinion survey was taken with local citizens in December that year, asking them how their views on nuclear energy changed after the accident. The survey findings indicate a grave fact that the citizens, who had lived in harmony with nuclear energy for the past 40 years, lost their confidence in this energy source because of the accident. The mishap also led to severer public opinion on nuclear energy at the nation-wide level, bringing forth a serious impediment to such projects as the siting of new nuclear power plants, the use of plutonium in light water reactors, intermediate storage of spent nuclear fuel, and the promotion of high-level radioactive waste disposal. Japan Atomic Energy Relations Organisation analysed the critical situation surrounding nuclear energy development in the country. Based on the analysis results, JAERO is now actively promoting an interactive, participatory program that meets people's needs, in an effort to regain public trust in nuclear energy. We, at JAERO, believe that greater importance should be attached to the offer of relevant information to the rising generation that will play a leading role in future society. The most typical of JAERO activities for junior and senior high school students include an Essay Content launched in 1968 in commemoration of Atomic Energy Day on October 26 every year and a Radiation Workshop established in 1978. Through more than 11,000 essays collected from junior and senior high school students across the county, this report discusses how these young people think of the JCO criticality accident and what they expect of nuclear energy development in the years ahead. Amid growing public concern with radiation following the JCO accident, the report also

Insights from the interim reliability evaluation program pertinent to reactor safety issues

International Nuclear Information System (INIS)

Carlson, D.D.

1983-01-01

The Interim Reliability Evaluation Program (IREP) consisted of concurrent probabilistic analyses of four operating nuclear power plants. This paper presents and integrated view of the results of the analyses drawing insights pertinent to reactor safety. The importance to risk of accident sequences initiated by transients and small loss-of-coolant accidents was confirmed. Support systems were found to contribute significantly to the sets of dominant accident sequences, either due to single failures which could disable one or more mitigating systems or due to their initiating plant transients. Human errors in response to accidents also were important risk contributors. Consideration of operator recovery actions influences accident sequence frequency estimates, the list of accident sequences dominating core melt, and the set of dominant risk contributors. Accidents involving station blackout, reactor coolant pump seal leaks and ruptures, and loss-of-coolant accidents requiring manual initiation of coolant injection were found to be risk significant

A preliminary study for the implementation of general accident management strategies

Energy Technology Data Exchange (ETDEWEB)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1998-12-31

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

A preliminary study for the implementation of general accident management strategies

Energy Technology Data Exchange (ETDEWEB)

Yang, Soo Hyung; Kim, Soo Hyung; Jeong, Young Hoon; Chang, Soon Heung [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of)

1997-12-31

To enhance the safety of nuclear power plants, implementation of accident management has been suggested as one of most important programs. Specially, accident management strategies are suggested as one of key elements considered in development of the accident management program. In this study, generally applicable accident management strategies to domestic nuclear power plants are identified through reviewing several accident management programs for the other countries and considering domestic conditions. Identified strategies are as follows; 1) Injection into the Reactor Coolant System, 2) Depressurize the Reactor Coolant System, 3) Depressurize the Steam Generator, 4) Injection into the Steam Generator, 5) Injection into the Containment, 6) Spray into the Containment, 7) Control Hydrogen in the Containment. In addition, the systems and instrumentation necessary for the implementation of each strategy are also investigated. 11 refs., 3 figs., 3 tabs. (Author)

Licensing decisions and safety research related to LMFBR accidents

International Nuclear Information System (INIS)

Denise, R.P.; Speis, T.P.; Kelber, C.N.; Curtis, R.T.

1977-01-01

The licensing approach which ensures adequate protection of the public health and safety against serious accidents is described. This paper describes the role of core melt and core disruptive accidents in the design, safety research, and licensing processes, using the Clinch River Breeder Reactor (CRBR) as a focal point. Major design attention is placed on the prevention of these accidents so that the probability of core melt accidents is reduced to a sufficiently low level that they are not treated as design basis accidents. Additional requirements are placed upon the design to further reduce residual risk. This licensing process is supported by a confirmatory research program designed to provide an independent basis for licensing judgements. It has as a goal the resolution of generic safety issues prior to the establishment of a commercial LMFBR industry. The program includes accident analysis, experiments in materials interactions, aerosol transport and system integrity and planning for new safety test facilities. The problems are approached in a multi-disciplinary functional manner that identifies key safety issues and centralizes efforts to resolve them. The near term objectives of the program support the licensing of the Clinch River Breeder Reactor (CRBR) and the proposed Prototype Large Breeder Reactor (PLBR). The long term objectives of the program support the licensing of commercial LMFBRs during the late 1980's and beyond. This safety research is designed to provide an independent basis for the licensing judgements which must be made by the Nuclear Regulatory Commission

CANDU safety under severe accidents

International Nuclear Information System (INIS)

Snell, V.G.; Howieson, J.Q.; Frescura, G.M.; King, F.; Rogers, J.T.; Tamm, H.

1988-01-01

The characteristics of the CANDU reactor relevant to severe accidents are set first by the inherent properties of the design, and second by the Canadian safety/licensing approach. Probabilistic safety assessment studies have been performed on operating CANDU plants, and on the 4 x 880 MW(e) Darlington station now under construction; furthermore a scoping risk assessment has been done for a CANDU 600 plant. They indicate that the summed severe core damage frequency is of the order of 5 x 10 -6 /year. CANDU nuclear plant designers and owner/operators share information and operational experience nationally and internationally through the CANDU Owners' Group (COG). The research program generally emphasizes the unique aspects of the CANDU concept, such as heat removal through the moderator, but it has also contributed significantly to areas generic to most power reactors such as hydrogen combustion, containment failure modes, fission product chemistry, and high temperature fuel behaviour. Abnormal plant operating procedures are aimed at first using event-specific emergency operating procedures, in cases where the event can be diagnosed. If this is not possible, generic procedures are followed to control Critical Safety Parameters and manage the accident. Similarly, the on-site contingency plans include a generic plan covering overall plant response strategy, and a specific plan covering each category of contingency

[Fatal occupational accidents in Lombardy].

Science.gov (United States)

Pianosi, G

1995-01-01

All fatal occupational accidents compensated in Lombardy from 1984 to 1989 were analyzed (1259 cases): significant differences between geographical distribution of fatal occupational accidents and workers were observed. Males accounted for about 95% of fatalities; an excess of cases was shown in both young and elderly workers. Death was the consequence of injuries involving most frequently the head, thorax and spinal cord. An excess of fatalities was observed in agriculture and, at a lower level, in manufacturing industries; small enterprises were involved in approximately 25% of fatalities occurring in the manufacturing industries and services. Employers were the victims of fatal accidents in 50% of cases in agriculture and in 70% of cases in craft industries. Construction, agriculture and transport accounted for about 50% of all fatalities. About 50% of fatal occupational accidents were related to vehicle use: the victim was the driver in the majority of cases, sometimes the victim was run over by a vehicle or fell from a vehicle. The results agree with some previous observations (e.g.: sex and age distribution; construction, agriculture and transport as working activities at high accident risk); but some original observations have emerged, in particular about the frequency of employers as victims and the role of vehicles in the genesis of fatal occupational accidents. If further studies confirm these latter observations, important developments could follow in preventive action design and implementation.

The nature of reactor accidents

International Nuclear Information System (INIS)

Domaratzki, Z.; Campbell, F.R.; Atchison, R.J.

1981-01-01

Reactor accidents are events which result in the release of radioactive material from a nuclear power plant due to the failure of one or more critical components of that plant. The failures, depending on their number and type, can result in releases whose consequences range from negligible to catastrophic. By way of examples, this paper describes four specific accidents which cover this range of consequence: failure of a reactor control system, loss of coolant, loss of coolant with impaired containment, and reactor core meltdown. For each a possible sequence of events and an estimate of the expected frequency are presented

SARTEMP2 - A computer program to calculate power and temperatures in a transport flask during a criticality accident

International Nuclear Information System (INIS)

Shaw, P.M.

1983-04-01

The computer code SARTEMP2, an extended version of the original SARTEMP program, which calculates the power and temperatures in a transport flask during a hypothetical criticality accident is described. The accident arises, it is assumed, during the refilling of the flask with water, bringing the system to delayed critical. As the water level continues to rise, reactivity is added causing the power to rise, and thus temperatures in the fuel, clad and water to increase. The point kinetics equations are coupled to the one-dimensional heat conduction equation. The model used, the method of solution of the equations and the input data required are given. (author)

Guide for licensing evaluations using CRAC2: A computer program for calculating reactor accident consequences

International Nuclear Information System (INIS)

White, J.E.; Roussin, R.W.; Gilpin, H.

1988-12-01

A version of the CRAC2 computer code applicable for use in analyses of consequences and risks of reactor accidents in case work for environmental statements has been implemented for use on the Nuclear Regulatory Commission Data General MV/8000 computer system. Input preparation is facilitated through the use of an interactive computer program which operates on an IBM personal computer. The resulting CRAC2 input deck is transmitted to the MV/8000 by using an error-free file transfer mechanism. To facilitate the use of CRAC2 at NRC, relevant background material on input requirements and model descriptions has been extracted from four reports - ''Calculations of Reactor Accident Consequences,'' Version 2, NUREG/CR-2326 (SAND81-1994) and ''CRAC2 Model Descriptions,'' NUREG/CR-2552 (SAND82-0342), ''CRAC Calculations for Accident Sections of Environmental Statements, '' NUREG/CR-2901 (SAND82-1693), and ''Sensitivity and Uncertainty Studies of the CRAC2 Computer Code,'' NUREG/CR-4038 (ORNL-6114). When this background information is combined with instructions on the input processor, this report provides a self-contained guide for preparing CRAC2 input data with a specific orientation toward applications on the MV/8000. 8 refs., 11 figs., 10 tabs

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.)

The nuclear accidents: Causes and consequences

International Nuclear Information System (INIS)

Rochd, M.

1988-01-01

The author discussed and compared the real causes of T.M.I. and Chernobyl accidents and cited their consequences. To better understand how these accidents occurred, a brief description of PWR type (reactor type of T.M.I.) and of RBMK type (reactor type of Chernobyl) has been presented. The author has also set out briefly the safety analysis objectives and the three barriers established to protect the public against the radiological consequences. To distinguish failures that cause severe accidents and to analyze them in details, it is necessary to classify the accidents. There are many ways to do it according to their initiator event, or to their frequency, or to their degree of gravity. The safety criteria adopted by nuclear industry have been explained. These criteria specify the limits of certain physical parameters that should not be exceeded in case of incidents or accidents. To compare the real causes of T.M.I. and Chernobyl accidents, the events that led to both have been presented. As observed the main common contributing factors in both cases are that the operators did not pay attention to warnings and signals that were available to them and that they were not trained to handle these accident sequences. The essential conclusions derived from these severe accidents are: -The improvement of operators competence contribute to reduce the accident risks; -The rapid and correct diagnosis of real conditions at each point of the accidents permits an appropriate behavior that would bring the plant to a stable state; -Competent technical teams have to intervene and to assist the operators in case of emergency; -Emergency plans and an international collaboration are necessary to limit the accident risks. 11 figs. (author)

Characteristics of severely damaged fuel from PBF tests and the TMI-2 accident

International Nuclear Information System (INIS)

Osetek, D.J.; Cook, B.A.; Dallman, R.J.; Broughton, J.M.

1986-01-01

As a result of the TMI-2 reactor accident, the US Nuclear Regulatory Commission initiated a research program to investigate phenomena associated with severe fuel damage accidents. This program is sponsored by several countries and includes in-pile and out-of-pile experiments, separate effects studies, and computer code development. The principal in-pile testing portion of the program includes four integral severe fuel damage (SFD) tests in the Power Burst Facility (PBF) at the Idaho National Engineering Laboratory (INEL). The INEL is also responsible for examining the damaged core in the Three Mile Island-Unit 2 (TMI-2) reactor, which offers the unique opportunity to directly compare the findings of an experimental program to those of an actual reactor accident. The principal core damage phenomena which can occur during a severe accident are discussed, and examples from the INEL research programs are used to illustrate the characteristics of these phenomena. The preliminary results of the programs are presented, and their impact on plant operability during severe accidents is discussed

HEATUP: a computer program for the thermal anaysis of a LOFC accident in an HTGR

International Nuclear Information System (INIS)

Siman-Tov, I.I.; Turner, W.D.

1976-11-01

The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short review of preliminary results related to the choice of thermal properties, boundary conditions and initial conditions; a full description of a typical three-dimensional R-Z model and a limited one of a two-dimensional RZ model. HEATUP's additional computations are presented together with the method of input preparation. The three-dimensional model of the Fulton Generating Station Loss of Forced Circulation accident is used as a sample problem. A complete presentation of the input data is made. Also, the computer printout of the sample problem input data and results are given

HEATUP: a computer program for the thermal anaysis of a LOFC accident in an HTGR

Energy Technology Data Exchange (ETDEWEB)

Siman-Tov, I.I.; Turner, W.D.

1976-11-01

The HEATUP code, a modification of the general, time-dependent, one-, two-, and three-dimensional program HEATING5, was designed for the thermal analysis of a Loss of Forced Circulation accident in a High Temperature Gas-Cooled Reactor. This report contains a description of the computational model which includes: a description of the basic problem; a short review of preliminary results related to the choice of thermal properties, boundary conditions and initial conditions; a full description of a typical three-dimensional R-Z model and a limited one of a two-dimensional RZ model. HEATUP's additional computations are presented together with the method of input preparation. The three-dimensional model of the Fulton Generating Station Loss of Forced Circulation accident is used as a sample problem. A complete presentation of the input data is made. Also, the computer printout of the sample problem input data and results are given.

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)

Off-site protective action selection for nuclear reactor accidents

International Nuclear Information System (INIS)

Weerakkody, S.D.

1986-01-01

A computer program based upon a model using a rational theoretical basis was developed to select appropriate off-site protective actions during nuclear reactor accidents. The special features of this program include (a) introduction of a precursor concept that uses the history of the accident progression to determine the spectrum of potential accident scenarios and estimates of the likelihoods of each accident scenario; (b) use of statistical decision theory and the concept of entropy of a spectrum to select the appropriate protective actions using either the minimax principle or the Bayes action method; and (c) introduction of methods to quantify evacuation travel risks. In order to illustrate the usefulness of the computer program, it was applied at three stages of the Three Mile Island accident scenario. Quantified non-radiological risks of evaluation have been used to establish dose thresholds below which evacuations are not justified. Using the Poisson analysis for evacuation risks and the absolute L-L BEIR model for radiation risk suggests 330 mrems as a reasonable value for this threshold. The usefulness of the program in developing a technical basis to select the size of the plume exposure pathway emergency planning zone (EPZ) is discussed. Quantified evacuation risks, cost, and the current rationale upon which the EPZ is based, justify an EPZ between 5-10 miles for WASH-1400 source-terms

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

Science.gov (United States)

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

2018-01-01

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

Accidents on ships in the Danish International Ship register

DEFF Research Database (Denmark)

Ádám, Balázs; Rasmussen, Hanna Barbara

to report accidents causing at least one day off work beyond the day of accident but the first source contains several accidents not fulfilling this criterion, too. Radio Medical is an independent service where all Danish ships may seek medical advice. The data sets were merged by identification number...... of our study is to describe trend of accidents and their contributing factors, with special focus on nationality, occurring in ships under Danish flag in the period 2010-2012. The study used two independent data sources, the Danish Maritime Authority and the Danish Radio Medical. It is mandatory...... to create a single database that has been studied by descriptive statistics and regression analysis. Findings show a stabilised number of accidents in the analysed period. The occurrence of accidents is influenced by nationality. There is a higher frequency of reported injuries found among Danish and other...

Injury protection and accident causation parameters for vulnerable road users based on German In-Depth Accident Study GIDAS.

Science.gov (United States)

Otte, Dietmar; Jänsch, Michael; Haasper, Carl

2012-01-01

Within a study of accident data from GIDAS (German In-Depth Accident Study), vulnerable road users are investigated regarding injury risk in traffic accidents. GIDAS is the largest in-depth accident study in Germany. Due to a well-defined sampling plan, representativeness with respect to the federal statistics is also guaranteed. A hierarchical system ACASS (Accident Causation Analysis with Seven Steps) was developed in GIDAS, describing the human causation factors in a chronological sequence. The accordingly classified causation factors - derived from the systematic of the analysis of human accident causes ("7 steps") - can be used to describe the influence of accident causes on the injury outcome. The bases of the study are accident documentations over ten years from 1999 to 2008 with 8204 vulnerable road users (VRU), of which 3 different groups were selected as pedestrians n=2041, motorcyclists n=2199 and bicyclists n=3964, and analyzed on collisions with cars and trucks as well as vulnerable road users alone. The paper will give a description of the injury pattern and injury mechanisms of accidents. The injury frequencies and severities are pointed out considering different types of VRU and protective measures of helmet and clothes of the human body. The impact points are demonstrated on the car, following to conclusion of protective measures on the vehicle. Existing standards of protection devices as well as interdisciplinary research, including accident and injury statistics, are described. With this paper, a summarization of the existing possibilities on protective measures for pedestrians, bicyclists and motorcyclists is given and discussed by comparison of all three groups of vulnerable road users. Also the relevance of special impact situations and accident causes mainly responsible for severe injuries are pointed out, given the new orientation of research for the avoidance and reduction of accident patterns. 2010 Elsevier Ltd. All rights reserved.

Early results from an experimental program to determine the behavior of containment piping penetration bellows subjected to severe accident conditions

International Nuclear Information System (INIS)

Lambert, L.D.; Parks, M.B.

1994-01-01

Containment piping penetration bellows are an integral part of the pressure boundary in steel containments in the United States (US). Their purpose is to minimize loading on the containment shell caused by differential movement between the piping and the containment. This differential movement is typically caused by thermal gradients generated during startup and shutdown of the reactor, but can be caused by earthquake, a loss-of-coolant accident (LOCA), or ''severe'' accidents. In the event of a severe accident, the bellows would be subjected to pressure, temperature, and deflection well beyond the design basis. Most bellows are installed such that they would be subjected to elevated internal pressure, elevated temperature, axial compression, and lateral deflection during a severe accident. A few bellows would be subjected to external pressure and axial elongation, as well as elevated temperature and lateral deflection. The purpose of this experimental program is to examine the potential for leakage of containment bellows during a severe accident. The test series subjects bellows to various levels and combinations of internal pressure, elevated temperature, axial compression or elongation, and lateral deformation. The experiments are being conducted in two parts. For Part 1, all bellows specimens are tested in ''like-new'' condition, without regard for the possible degrading effect of corrosion that has been observed in some containment piping bellows in the US Part I testing, which included 13 bellows tests, has been completed. The second part of the experimental program, in which bellows are subjected to simulated corrosive environments prior to testing, has just just begun. The Part I experiments have shown that bellows in ''like-new'' condition can withstand elevated temperatures and pressures along with large deformations before leaking. In most cases, the like-new bellows were fully compressed without developing any leakage

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

International Nuclear Information System (INIS)

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

1992-01-01

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

The role of OSHA violations in serious workplace accidents.

Science.gov (United States)

Mendeloff, J

1984-05-01

California accident investigations for 1976 show that violations of the Occupational Safety and Health Administration's safety standards were a contributing factor in 13% to 19% of the 645 deaths reported to the workers' compensation program during that year. However, a panel of safety engineers judged that only about 50% of these violations could have been detected if an inspector had visited the day before the accident. These findings indicate that the potential gains from stronger enforcement of current standards are limited but not insignificant. The likelihood that a violation contributed to a serious accident varied considerably among accident types, industries, and size classes of plants. These findings can be used to increase the efficiency and effectiveness of the OSHA program by means of better targeting of inspections and accident investigations, more intelligent assessment of which violations should be penalized most heavily, and the provision of information to employers and workers about which violations are most consequential.

Severe accident risks: An assessment for five US nuclear power plants: Appendices A, B, and C

International Nuclear Information System (INIS)

1990-12-01

This report summarizes an assessment of the risks from severe accidents in five commercial nuclear power plants in the United States. These risks are measured in a number of ways, including: the estimated frequencies of core damage accidents from internally initiated accidents and externally initiated accidents for two or the plants; the performance of containment structures under severe accident loadings; the potential magnitude of radionuclide release and offsite consequences of such accidents; and the overall risk (the product of accident frequencies and consequences). Supporting this summary report are a large number of reports written under contract to NRC that provide the detailed discussion of the methods used and results obtained in these risk studies. Volume 2 of this report contains three appendices, providing greater detail on the methods used, an example risk calculation, and more detailed discussion of particular technical issues found important in the risk studies

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.

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Work time control, sleep & accident risk: A prospective cohort study.

Science.gov (United States)

Tucker, Philip; Albrecht, Sophie; Kecklund, Göran; Beckers, Debby G J; Leineweber, Constanze

We examined whether the beneficial impact of work time control (WTC) on sleep leads to lower accident risk, using data from a nationally representative survey conducted in Sweden. Logistic regressions examined WTC in 2010 and 2012 as predictors of accidents occurring in the subsequent 2 years (N = 4840 and 4337, respectively). Sleep disturbance and frequency of short sleeps in 2012 were examined as potential mediators of the associations between WTC in 2010 and subsequent accidents as reported in 2014 (N = 3636). All analyses adjusted for age, sex, education, occupational category, weekly work hours, shift work status, job control and perceived accident risk at work. In both waves, overall WTC was inversely associated with accidents (p = 0.048 and p = 0.038, respectively). Analyses of the sub-dimensions of WTC indicated that Control over Daily Hours (influence over start and finish times, and over length of shift) did not predict accidents in either wave, while Control over Time-off (CoT; influence over taking breaks, running private errands during work and taking paid leave) predicted fewer accidents in both waves (p = 0.013 and p = 0.010). Sleep disturbance in 2012 mediated associations between WTC/CoT in 2010 and accidents in 2014, although effects' sizes were small (effectWTC = -0.006, 95% confidence interval [CI] = -0.018 to -0.001; effectCoT = -0.009, 95%CI = -0.022 to -0.001; unstandardized coefficients), with the indirect effects of sleep disturbance accounting for less than 5% of the total direct and indirect effects. Frequency of short sleeps was not a significant mediator. WTC reduces the risk of subsequently being involved in an accident, although sleep may not be a strong component of the mechanism underlying this association.

ARAMIS project: A comprehensive methodology for the identification of reference accident scenarios in process industries

International Nuclear Information System (INIS)

Delvosalle, Christian; Fievez, Cecile; Pipart, Aurore; Debray, Bruno

2006-01-01

In the frame of the Accidental Risk Assessment Methodology for Industries (ARAMIS) project, this paper aims at presenting the work carried out in the part of the project devoted to the definition of accident scenarios. This topic is a key-point in risk assessment and serves as basis for the whole risk quantification. The first result of the work is the building of a methodology for the identification of major accident hazards (MIMAH), which is carried out with the development of generic fault and event trees based on a typology of equipment and substances. The term 'major accidents' must be understood as the worst accidents likely to occur on the equipment, assuming that no safety systems are installed. A second methodology, called methodology for the identification of reference accident scenarios (MIRAS) takes into account the influence of safety systems on both the frequencies and possible consequences of accidents. This methodology leads to identify more realistic accident scenarios. The reference accident scenarios are chosen with the help of a tool called 'risk matrix', crossing the frequency and the consequences of accidents. This paper presents both methodologies and an application on an ethylene oxide storage

Status of the French R/D program on the severe accident issue to develop Gen IV SFRs - 15373

International Nuclear Information System (INIS)

Serre, F.; Bertrand, F.; Journeau, C.; Suteau, C.; Verwaede, D.; Schmitt, D.; Farges, B.

2015-01-01

The ASTRID reactor (Advanced Sodium Technological Reactor for Industrial Demonstration) is a technological demonstrator designed by the CEA with its industrial partners, with very high levels of requirements. In the ASTRID project, the safety objectives are to prevent core melting, in particular by the development of an innovative core with complementary safety prevention devices, and to enhance the reactor resistance to severe accident by design. To mitigate the consequences of hypothetical core melting situations, specific dispositions or mitigation devices will be added to the core and to the reactor. It is also required to provide a robust safety demonstration (with high level of confidence). Therefore a new approach for severe accident issue has been defined: to the well-known 'lines of defense' method, a 'lines of mitigation' method is added. To meet these ASTRID, or future SFR, requirements, a large R/D program was launched in the Severe Accident domain, with a large number of partners. This paper will present the status of the CEA R/D related to the SFR Severe Accident issue, the collaboration framework (with industrial partners and R/D foreign organizations), and the future R/D plans to support the ASTRID project and possible developments for future Gen IV commercial SFR. (authors)

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

Head and face injuries and helmet use among injured motorcyclists with road accidents in Isfahan, Iran

Directory of Open Access Journals (Sweden)

Azam Dadkhah

2015-10-01

Full Text Available BACKGROUND AND AIM: The study aimed to assess the frequency of head and face injuries in motorcyclists who had an accident and to find out the relationship between helmet use and frequency of these injuries. METHODS: A cross-sectional study with multi-stage sampling method provides data on the injured motorcyclists with road accidents. Data came from a registration form which has documented information of each injured person who had a road accident and hospitalized in the biggest hospital of Isfahan University of Medical Sciences, Iran (Al-Zahra. All the registration forms were surveyed for hospitalization period, treatment costs, severity of injury, and date of accident during 2010 (n = 1626. Later, among the list of injured motorcyclists during the last 3 months of the registration form, 125 cases were randomly selected and interviewed by phone regarding occurrence of the head and face injuries and whether wearing helmet during the accident. Confidence intervals (CI, Chi-square, and Phi and Cramer’s correlation coefficient were applied. The ethical approval was provided. RESULTS: Accident by motorcycle was 31.0% of all road accidents. The frequency of motorcycle accidents was higher in the autumn and among 21-25 year olds. The mean period of hospitalization was 4.3 days and the mean of hospital costs was about 9000000 Rials [about 8200 United States dollar (USD, in 2010]. Of motorcyclist, 35.0% reported they were helmeted when they had the accident. The frequency of head and face injuries was 51.0% among all the injured motorcyclists, 22.0% and 78.0% among the helmeted and non-helmeted motorcyclists, respectively (P = 0.009, r = -0.267. CONCLUSION: Motorcycle accidents comprise a large number of road accidents and cause substantial morbidity and financial impact for the community members. Head and face injuries are the most common trauma in motorcyclists, and the injury rate is higher among non-helmeted motorcyclists.

Generic implications of the Chernobyl accident

International Nuclear Information System (INIS)

Sege, G.

1989-01-01

The US Nuclear Regulatory Commission (NRC) staff's assessment of the generic implications of the Chernobyl accident led to the conclusion that no immediate changes in the NRC's regulations regarding design or operation of US commercial reactors are needed. However, further consideration of certain issues was recommended. This paper discusses those issues and the studies being addressed to them. Although 24 tasks relating to light water reactor issues are identified in the Chernobyl follow-up research program, only four are new initiatives originating from Chernobyl implications. The remainder are limited modifications of ongoing programs designed to ensure that those programs duly reflect any lessons that may be drawn from the Chernobyl experience. The four new study tasks discussed include a study of reactivity transients, to reconfirm or bring into question the adequacy of potential reactivity accident sequences hitherto selected as a basis for design approvals; analysis of risk at low power and shutdown; a study of procedure violations; and a review of current NRC testing requirements for balance of benefits and risks. Also discussed, briefly, are adjustments to ongoing studies in the areas of operational controls, design, containment, emergency planning, and severe accident phenomena

Evaluation of Safety Programs with Respect to the Causes of General Aviation Accidents. Volume I. Technical Report,

Science.gov (United States)

1980-05-01

65 Physical Impairment 66 Spatial disorientation. 67 Psychological condition. 71 Misused or failed to use flaps. 74 Left aircraft unattended, engine...ARTS III - (Software) (1975) 203 Weather Radar Display System (ASR - 57) 204 ATARS - Automated Terminal Area Radar Service (1974) 205 Instrument Landing...Generated Trauma, Pathological and Psychological Dysfunction accident causes. Collectively, the distribution of safety programs throughout the fault

Severe accident testing of electrical penetration assemblies

International Nuclear Information System (INIS)

Clauss, D.B.

1989-11-01

This report describes the results of tests conducted on three different designs of full-size electrical penetration assemblies (EPAs) that are used in the containment buildings of nuclear power plants. The objective of the tests was to evaluate the behavior of the EPAs under simulated severe accident conditions using steam at elevated temperature and pressure. Leakage, temperature, and cable insulation resistance were monitored throughout the tests. Nuclear-qualified EPAs were produced from D. G. O'Brien, Westinghouse, and Conax. Severe-accident-sequence analysis was used to generate the severe accident conditions (SAC) for a large dry pressurized-water reactor (PWR), a boiling-water reactor (BWR) Mark I drywell, and a BWR Mark III wetwell. Based on a survey conducted by Sandia, each EPA was matched with the severe accident conditions for a specific reactor type. This included the type of containment that a particular EPA design was used in most frequently. Thus, the D. G. O'Brien EPA was chosen for the PWR SAC test, the Westinghouse was chosen for the Mark III test, and the Conax was chosen for the Mark I test. The EPAs were radiation and thermal aged to simulate the effects of a 40-year service life and loss-of-coolant accident (LOCA) before the SAC tests were conducted. The design, test preparations, conduct of the severe accident test, experimental results, posttest observations, and conclusions about the integrity and electrical performance of each EPA tested in this program are described in this report. In general, the leak integrity of the EPAs tested in this program was not compromised by severe accident loads. However, there was significant degradation in the insulation resistance of the cables, which could affect the electrical performance of equipment and devices inside containment at some point during the progression of a severe accident. 10 refs., 165 figs., 16 tabs

Biological effects in natural populations of small rodents in radiocontaminated areas. The frequency of bone marrow polyploid cells in bank voles in different years following the Chernobyl accident

International Nuclear Information System (INIS)

Ryabokon', N.I.

1999-01-01

On the basis of metaphase analysis results the peculiarities of dynamics of genome mutation frequency (polyploid cells) were studied in bone marrow of bank voles inhibiting the areas with different contamination level due to the Chernobyl accident (8-1526 kBq/m 2 for 137 Cs) in 1986-1991. Unexpectedly high frequencies of polyploid cells exceeding the pre-accidental level by factor of 10 1 -10 3 were recorded in all populations studied. Relationship between the frequency of parameter studied and the concentration of radionuclides incorporated in animal carcasses was proved. Statistically significant rise in the frequency of genome mutations with the time was revealed up to 1991, i.e. approximately to 12-th post-accidental animal generation [ru

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)

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

Group unified accident reporting database (GUARD)

International Nuclear Information System (INIS)

Koene, W.; Waterfall, K.W.

1991-01-01

Significant advances have been made in recent years in enhancing the standard of safety within Shell Companies, such that safety has now been raised to a status equal to other primary business objectives. It is widely accepted that accident prevention is part of good business practice, and that a safe operation is normally an efficient operation. Safety programmes are being widely implemented which involve all employees from top management right down to the workforce including the contract staff, and the benefits are being realized. The effectiveness of any safety programme, however, must be continuously monitored, and in this respect injury and accident statistics play an important role as a prime indicator of safety performance. Statistics form part of the safety management process indicating the success of the safety programmes being implemented, and highlighting areas of weakness. Statistical information relating to the number and frequency of accidents, significant as it is, tells us little about how the accidents occur, or about how to improve the intrinsic safety of the operations. More detailed information on accident causes and lessons derived from the investigation of non-injurious accidents and near-misses is required for this, and for the setting of appropriate remedial actions. This paper concentrates on the feedback from accidents which have already occurred. This feedback plays a vital role as an indicator of safety performance upon which to judge the effectiveness of safety programmes, and also to provide important information relating to the immediate and underlying causes of accidents. To meet these requirements, however, a system for recording analyzing and communicating safety data is essential

Severe Accident Test Station Activity Report

Energy Technology Data Exchange (ETDEWEB)

Pint, Bruce A [ORNL; Terrani, Kurt A [ORNL

2015-06-01

Enhancing safety margins in light water reactor (LWR) severe accidents is currently the focus of a number of international R&D programs. The current UO2/Zr-based alloy fuel system is particularly susceptible since the Zr-based cladding experiences rapid oxidation kinetics in steam at elevated temperatures. Therefore, alternative cladding materials that offer slower oxidation kinetics and a smaller enthalpy of oxidation can significantly reduce the rate of heat and hydrogen generation in the core during a coolant-limited severe accident. In the U.S. program, the high temperature steam oxidation performance of accident tolerant fuel (ATF) cladding solutions has been evaluated in the Severe Accident Test Station (SATS) at Oak Ridge National Laboratory (ORNL) since 2012. This report summarizes the capabilities of the SATS and provides an overview of the oxidation kinetics of several candidate cladding materials. A suggested baseline for evaluating ATF candidates is a two order of magnitude reduction in the steam oxidation resistance above 1000ºC compared to Zr-based alloys. The ATF candidates are categorized based on the protective external oxide or scale that forms during exposure to steam at high temperature: chromia, alumina, and silica. Comparisons are made to literature and SATS data for Zr-based alloys and other less-protective materials.

National practices in relation to severe accidents

International Nuclear Information System (INIS)

Soda, Kunihisa

1989-01-01

After the accidents at Three Mile Island and Chernobyl, many studies have been carried out on severe accidents by various organizations including IAEA and OECD/CSNI. In the present article, measures taken in different countries against severe accidents are outlined based on the results of these studies. In Sweden, policies for the management of a severe accident and reduction in the release of radioactive materials were established based on reports issued by the Atomic Energy Committee, which was set up after the Three Mile Island accident. The current policies require that filter vents be provided where necessary. France, following Sweden, adopted the use of filter vents. Operation procedures to be followed in the event of a severe accident have been established in the nation. The measures against severe accidents adopted in West Germany mainly focus on the weakening of the effects of accidents, and are not covered by the design standards. The use of filter vents are also required in Finland and Switzerland. In the U.S., a program for individual plant examination will be implemented over the three-year period beginning in 1989. Studies on measures against severe accidents seem to be performed also in the Soviet Union. (N.K.)

Probability safety assessment of LOOP accident to molten salt reactor

International Nuclear Information System (INIS)

Mei Mudan; Shao Shiwei; Yu Zhizhen; Chen Kun; Zuo Jiaxu

2013-01-01

Background: Loss of offsite power (LOOP) is a possible accident to any type of reactor, and this accident can reflect the main idea of reactor safety design. Therefore, it is very important to conduct a study on probabilistic safety assessment (PSA) of the molten salt reactor that is under LOOP circumstance. Purpose: The aim is to calculate the release frequency of molten salt radioactive material to the core caused by LOOP, and find out the biggest contributor to causing the radioactive release frequency. Methods: We carried out the PSA analysis of the LOOP using the PSA process risk spectrum, and assumed that the primary circuit had no valve and equipment reliability data based on the existing mature power plant equipment reliability data. Results: Through the PSA analysis, we got the accident sequences of the release of radioactive material to the core caused by LOOP and its frequency. The results show that the release frequency of molten salt radioactive material to the core caused by LOOP is about 2×10 -11 /(reactor ·year), which is far below that of the AP1000 LOOP. In addition, through the quantitative analysis, we obtained the point estimation and interval estimation of uncertainty analysis, and found that the biggest contributor to cause the release frequency of radioactive material to the core is the reactor cavity cooling function failure. Conclusion: This study provides effective help for the design and improvement of the following molten salt reactor system. (authors)

Package of programs for calculating accidents involving melting of the materials in a fast-reactor vessel

International Nuclear Information System (INIS)

Vlasichev, G.N.

1994-01-01

Methods for calculating one-dimensional nonstationary temperature distribution in a system of physically coupled materials are described. Six computer programs developed for calculating accident processes for fast reactor core melt are described in the article. The methods and computer programs take into account melting, solidification, and, in some cases, vaporization of materials. The programs perform calculations for heterogeneous systems consisting of materials with arbitrary but constant composition and heat transfer conditions at material boundaries. Additional modules provide calculations of specific conditions of heat transfer between materials, the change in these conditions and configuration of the materials as a result of coolant boiling, melting and movement of the fuel and structural materials, temperature dependences of thermophysical properties of the materials, and heat release in the fuel. 11 refs., 3 figs

Review of U.S. Army Unmanned Aerial Systems Accident Reports: Analysis of Human Error Contributions

Science.gov (United States)

2018-03-20

within report documents. The information presented was obtained through a request to use the U.S. Army Combat Readiness Center’s Risk Management ...controlled flight into terrain (13 accidents), fueling errors by improper techniques (7 accidents), and a variety of maintenance errors (10 accidents). The...and 9 of the 10 maintenance accidents. Table 4. Frequencies Based on Source of Human Error Human error source Presence Poor Planning

Scoping Study Investigating PWR Instrumentation during a Severe Accident Scenario

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L. [Rempe and Associates, LLC, Idaho Falls, ID (United States); Knudson, D. L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Lutz, R. J. [Lutz Nuclear Safety Consultant, LLC, Asheville, NC (United States)

2015-09-01

The accidents at the Three Mile Island Unit 2 (TMI-2) and Fukushima Daiichi Units 1, 2, and 3 nuclear power plants demonstrate the critical importance of accurate, relevant, and timely information on the status of reactor systems during a severe accident. These events also highlight the critical importance of understanding and focusing on the key elements of system status information in an environment where operators may be overwhelmed with superfluous and sometimes conflicting data. While progress in these areas has been made since TMI-2, the events at Fukushima suggests that there may still be a potential need to ensure that critical plant information is available to plant operators. Recognizing the significant technical and economic challenges associated with plant modifications, it is important to focus on instrumentation that can address these information critical needs. As part of a program initiated by the Department of Energy, Office of Nuclear Energy (DOE-NE), a scoping effort was initiated to assess critical information needs identified for severe accident management and mitigation in commercial Light Water Reactors (LWRs), to quantify the environment instruments monitoring this data would have to survive, and to identify gaps where predicted environments exceed instrumentation qualification envelop (QE) limits. Results from the Pressurized Water Reactor (PWR) scoping evaluations are documented in this report. The PWR evaluations were limited in this scoping evaluation to quantifying the environmental conditions for an unmitigated Short-Term Station BlackOut (STSBO) sequence in one unit at the Surry nuclear power station. Results were obtained using the MELCOR models developed for the US Nuclear Regulatory Commission (NRC)-sponsored State of the Art Consequence Assessment (SOARCA) program project. Results from this scoping evaluation indicate that some instrumentation identified to provide critical information would be exposed to conditions that

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

OpenAIRE

Mehran Amiri; Abdollah Ardeshir; Mohammad Hossein Fazel Zarandi

2015-01-01

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 d...

Response to the Chernobyl accident in Japan

International Nuclear Information System (INIS)

Anon.

1986-01-01

The worst nuclear accident in history happened at No.4 unit of the Chernobyl Atomic Power Station in USSR. Since the Chernobyl accident, a number of measures have been introduced in many countries, including the reconsideration of programs for construction and operation of nuclear power plants. In Japan, the press and television first reported the accident on April 29. The next day, all the relevant governmental agencies began to collect and analyze information in order to prepare possible countermeasures. The Nuclear Safety Commission issued a statement covering three points: 1) the radioactive substances released by the accident will have virtually no influence on the health of people in Japan, 2) a Special Committee on the Chernobyl Atomic Power Station Accident will be established, and 3) the Soviet government must provide all detailed information about the accident as soon as it is available. On April 30, the Committee on Radioactivity decided to increase radioactivity observations by the Science and Technology Agency, the Defence Agency, and the Meteorological Agency. On the same day, the Ministry of International Trade and Industry set up a survey committee for the Chernobyl accident with the responsibility of collecting and analyzing information about the accident. A review is also made in this article as to how the Japanese media reported the accident and how people reacted on reading the newspapers and watching TV on the accident. (Nogami, K.)

Ergonomic study of biorhythm effect on the 62 occurrence of human errors and accidents in automobile manufacturing industry

Directory of Open Access Journals (Sweden)

2012-03-01

Conclusion: This study showed that the frequency of accidents in critical days and negative section of physical cycle was more than expected. Also the frequency of accidents in critical days and negative section of emotional and intellectual cycle was less than expected. Due to the physical nature of the work activities in the automobile manufacturing industry can be stated that the study showed that in physical work activities, frequency of accidents in critical days and negative section of physical cycle in which the person is not physically ready to do the job was more than expected.

Accident dynamics of LR-0 reactor

International Nuclear Information System (INIS)

Vorisek, M.; Tinka, I.

1981-01-01

The results are given of calculating the accident dynamics of the LR-0 light water experimental zero power reactor. Calculations of the time dependence of power, the total released energy, the temperature of fuel and its cladding were made using program FATRAP for different values of the total inserted reactivity. Using the results, an analysis is made of hypothetic accident states of the LR-0 reactor. The results are shown graphically. (J.B.)

An overview of selected severe accident research and applications

International Nuclear Information System (INIS)

Hammersley, R.J.; Henry, R.E.

2004-01-01

Severe accident research is being conducted world wide by industry organizations, utilities, and regulatory agencies. As this research is disseminated, it is being applied by utilities when they perform their Individual Plant Examinations (IPEs) and consider the preparation of Accident Management programs. The research is associated with phenomenological assessments of containment challenges and associated uncertainties, severe accident codes and analysis tools, systematic evaluation processes, and accident management planning. The continued advancement of this research and its applications will significantly contribute to the enhanced safety and operation of nuclear power plants. (author)

Severe accident testing of a personnel airlock

International Nuclear Information System (INIS)

Clauss, D.B.; Parks, M.B.; Julien, J.T.; Peters, S.W.

1988-01-01

Sandia National Laboratories (Sandia) is investigating the leakage potential of mechanical penetrations as part of a research program on containment integrity under severe accident loads for the U.S. Nuclear Regulatory Commission (NRC). Barnes et al. (1984) and Shackelford et al. (1985) identified leakage from personnel airlocks as an important failure mode of containments subject to severe accident loads. However, these studies were based on relatively simple analysis methods. The complex structural interaction between the door, gasket, and bulkhead in personnel airlocks makes analytical evaluation of leakage difficult. In order to provide data to validate methods for evaluating the leakage potential, a full-size personnel airlock was subject to simulated severe accident loads consisting of pressure and temperature up to 300 psig and 800 degrees F. The test was conducted at Chicago Bridge and Iron under contract to Sandia. The authors provide a detailed report on the test program

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

International Nuclear Information System (INIS)

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

1992-01-01

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

SAMEX: A severe accident management support expert

International Nuclear Information System (INIS)

Park, Soo-Yong; Ahn, Kwang-Il

2010-01-01

A decision support system for use in a severe accident management following an incident at a nuclear power plant is being developed which is aided by a severe accident risk database module and a severe accident management simulation module. The severe accident management support expert (SAMEX) system can provide the various types of diagnostic and predictive assistance based on the real-time plant specific safety parameters. It consists of four major modules as sub-systems: (a) severe accident risk data base module (SARDB), (b) risk-informed severe accident risk data base management module (RI-SARD), (c) severe accident management simulation module (SAMS), and (d) on-line severe accident management guidance module (on-line SAMG). The modules are integrated into a code package that executes within a WINDOWS XP operating environment, using extensive user friendly graphics control. In Korea, the integrated approach of the decision support system is being carried out under the nuclear R and D program planned by the Korean Ministry of Education, Science and Technology (MEST). An objective of the project is to develop the support system which can show a theoretical possibility. If the system is feasible, the project team will recommend the radiation protection technical support center of a national regulatory body to implement a plant specific system, which is applicable to a real accident, for the purpose of immediate and various diagnosis based on the given plant status information and of prediction of an expected accident progression under a severe accident situation.

Simulation of a programmed frequency shift near extraction from the Fermilab Booster

International Nuclear Information System (INIS)

Lucas, P.; Kerns, Q.

1987-03-01

The longitudinal phase space program ESME has been used to simulate the effects of a linear shift in RF frequency away from that appropriate for the accelerator guide field. This shift takes place in the new Booster low level RF and is used to position the particle bunches in Main Ring buckets in a reproducible fashion. Shifts in frequency are found to generate synchrotron oscillations; however, the simulations show that these can be reduced to acceptable levels by introduction of jumps in RF phase preceding the programmed frequency changes. Lowering the RF voltage near extraction from the Booster, a desirable operational feature, has also been investigated

The frequency of chromatide aberrations as a function of radiation dose estimated by the number of dicentrics found by the cytogenetic analysis of lymphocytes in subjects affected by the Chernobyl accident

International Nuclear Information System (INIS)

Nugis, V.Yu.; Chirkov, A.A.

1990-01-01

A study was made of the frequency of chromatide aberrations in lymphocyte culture of subjects affected by the Chernobyl accident as a function of dose estimated by the incidence of dicentrics. The average number of chromatide aberrations was nearly the same within the dose range from 0 to 5 Gy exhibiting a tendency towards growth with dose. A high individual variability of the chromatide aberration frequency was observed

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

Occupational Accidents: A Perspective of Pakistan Construction Industry

Directory of Open Access Journals (Sweden)

Tauha Hussain Ali

2014-07-01

Full Text Available It has been observed that the construction industry is one of the notorious industry having higher rate of fatalities and injuries. Resulting in higher financial losses and work hour losses, which are normally faced by this industry due to occuptional accidents. Construction industry has the highest occupational accidents rate recorded throughout the world after agriculture industry. The construction work site is often a busy place having an incredibly high account of activities taking place, where everyone is moving in frenzy having particular task assigned. In such an environment, occupational accidents do occur. This paper gives information about different types of occupational accidents & their causes in the construction industry of Pakistan. A survey has been carried out to identify the types of occupational accidents often occur at construction site. The impact of each occupational accident has also been identified. The input from the different stakeholders involved on the work site was analyzed using RIW (Relative Importance Weight method. The findings of this research show that ?fall from elevation, electrocution from building power and snake bite? are the frequent occupational accidents occur within the work site where as ?fall from elevation, struck by, snake bite and electrocution from faulty tool? are the occupational accident with high impact within the construction industry of Pakistan. The results also shows the final ranking of the accidents based on higher frequency and higher impact. Poor Management, Human Element and Poor Site Condition are found as the root causes leading to such occupational accidents. Hence, this paper

NEK AOV program

International Nuclear Information System (INIS)

Fable, D.; Andric, A.; Prah, M.; Habinc, M.

2002-01-01

Air Operated Valves (AOVs) are used in a wide variety of applications in nuclear power plants. Some AOVs perform important functions in safety and non-safety related systems which could affect initiating event frequencies, accident mitigation and radiological releases. A review of lessons learned from nuclear plant MOV programs and pilot AOV programs indicates that AOV performance can be enhanced via improvements in valve and actuator sizing, setup, testing and maintenance. Nuklearna elektrarna Krsko has decided to develop an AOV program on the basis of Joint Owners. Group (JOG) AOV Program, which identifies nine key elements for successful implementation. The first step in establishing an AOV Program is to identify and categorize the plant AOVs for evaluation. Performed categorization in NEK has resulted with Preliminary AOV Scope. Important part of NEK AOV Program is NEK AOV Database. For the time being, basic structure with tables, queries, forms, reports and relationships has been created. Further development will be coordinated with development of NEK AOV Program Methodology.(author)

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.

Identification of the operating crew's information needs for accident management

International Nuclear Information System (INIS)

Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

1988-01-01

While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence

Accidents in the greenhouse-construction industry of SE Spain.

Science.gov (United States)

Pérez-Alonso, José; Carreño-Ortega, Angel; Vázquez-Cabrera, Fernando J; Callejón-Ferre, Angel Jesús

2012-01-01

This work analyses the labour accidents in the greenhouse-construction industry of SE Spain for the period 1999-2007 through a sample of 180 accident reports. The accidents were characterised by studying 5 variables in order to know the day of the week in which the accident occurred, the hour of the day of the accident, type of accident, the region of Spain in which the accident happened, and the resulting injury. The data characterising the accidents were submitted to a descriptive multiple-correspondence analysis. The incidence of accidents in the greenhouse-construction industry presented a high mean value of 15133.7 per 100,000 workers per year. The days with the greatest incidence of accidents were Thursday and Monday, while the period of greatest number of accidents occurred in the first 4h of the workday. No significant correspondence was found between the day of the week, the hour of the day, or any of the other 3 variables studied. The types of accidents with most frequency were: cuts, punctures, contact with hard or rough material, overexertion, and falls from one level to another. The most affected parts of the anatomy were the eyes, thorax, back, sides, lower legs, and feet. The most common types of injury were bone fractures, twists and sprains, distended muscles, contusions, and being crushed. The study calls attention to the high number of accidents at work, which needs to be corrected by fulfilment of safety regulations at work, on the part of the company. Finally, recommendations are made to correct this situation of high number of accidents at work. Copyright © 2011 Elsevier Ltd and The Ergonomics Society. All rights reserved.

Fifteen years of the cuban program, with children from areas affected by the Chernobyl accident

International Nuclear Information System (INIS)

Garcia, Omar; Medina, Julio

2005-01-01

The Cuban Chernobyl Program arrived to the 15th anniversary in March 2005. This program was designed to offers specialised medical attention and to develop a rehabilitation plan with children from areas affected by the Chernobyl accident. More than 21 500 children and adults had been assisted in the program up to the moment, with a significant set of medical procedures done. Dosimetric and biomedic research had been also carried out as part of the program. The most significant medical attention activities include the treatment of children with haematological disorders, including 120 leukaemia, bone marrow transplants, and the treatment of endocrine and cancer diseases. The dosimetric studies allow made a data base with information on internal 137Cs contamination, internal, external and total doses, children living location, and its contamination by 137Cs, and other significant information for radiological impact evaluation in more than 8000 children. The behaviour of all the medical information of the program in relation to the contamination of the land and the internal contamination of the children was analysed using this database. The program has accumulated an experience of interest for physicians, psychologists and in general persons interested in Chernobyl consequences. This paper present a summary of the main results obtained in the program

SIMMER Program: its accomplishments

International Nuclear Information System (INIS)

Smith, L.L.; Bell, C.R.; Bohl, W.R.; Luck, L.B.; Wehner, T.R.; DeVault, G.P.; Parker, F.R.

1985-01-01

SIMMER's mechanistic framework has contributed to substantial progress in understanding severe LMFBR accidents, the important accident processes, the process environments, and the need for consistency in assessing the total accident sequence. A comprehensive program of model and code development, experiment analyses, separate-effects studies, and whole-core accident calculations has resolved issues in transition-phase, behavior and showed convincingly that Clinch River Breeder Reactor accident energetics were within design requirements. The current international program continues progress toward improving our understanding of severe accidents in LMFBRs. 17 refs

Examining accident reports involving autonomous vehicles in California.

Directory of Open Access Journals (Sweden)

Francesca M Favarò

Full Text Available Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017. The data provides important information on autonomous vehicles accidents' dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama.

Examining accident reports involving autonomous vehicles in California.

Science.gov (United States)

Favarò, Francesca M; Nader, Nazanin; Eurich, Sky O; Tripp, Michelle; Varadaraju, Naresh

2017-01-01

Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017). The data provides important information on autonomous vehicles accidents' dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama.

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.

Application of precursor methodology in initiating frequency estimates

International Nuclear Information System (INIS)

Kohut, P.; Fitzpatrick, R.G.

1991-01-01

The precursor methodology developed in recent years provides a consistent technique to identify important accident sequence precursors. It relies on operational events (extracting information from actual experience) and infers core damage scenarios based on expected safety system responses. The ranking or categorization of each precursor is determined by considering the full spectrum of potential core damage sequences. The methodology estimates the frequency of severe core damage based on the approach suggested by Apostolakis and Mosleh, which may lead to a potential overestimation of the severe-accident sequence frequency due to the inherent dependencies between the safety systems and the initiating events. The methodology is an encompassing attempt to incorporate most of the operating information available from nuclear power plants and is an attractive tool from the point of view of risk management. In this paper, a further extension of this methodology is discussed with regard to the treatment of initiating frequency of the accident sequences

Occupational accidents: a perspective of pakistan construction industry

International Nuclear Information System (INIS)

Ali, T.H.; Khahro, S.H.; Memon, F.A.

2014-01-01

It has been observed that the construction industry is one of the notorious industry having higher rate of facilities and injuries. Resulting in higher financial losses and work hour losses, which are normally faced by this industry due to occupational accidents. Construction industry has the highest occupational accidents rate recorded throughout the world after agriculture industry. The construction work site is often a busy place having an incredibly high account of activities taking place, where everyone is moving in frenzy having particular task assigned. In such an environment, occupational accidents do occur. This paper gives information about different types of occupational accidents and their causes in the construction industry of Pakistan. A survey has been carried out to identify the types of occupational accidents often occur at construction site. The impact of each occupational accident has also been identified. The input from the different stakeholders involved on the work site was analyzed using RIW (Relative Importance Weight) method. The findings of this research show that fall from elevation, electrocution from building power and snake bite are the frequent occupational accidents occur within the work site where as fall from elevation, struck by, snake bite and electrocution from faulty tool are the occupational accident with high impact within the construction industry of Pakistan. The results also shows the final ranking of the accidents based on higher frequency and higher impact. Poor Management, Human Element and Poor Site Condition are found as the root causes leading to such occupational accidents. Hence, this paper identify that what type of occupational accidents occur at the work place in construction industry of pakistan, in order to develop the corrective actions which should be adequate enough to prevent the re-occurrence of such accidents at work site. (author)

Jose Cabrera NPP severe accident management activities

International Nuclear Information System (INIS)

Blanco, J.; Almeida, P.; Saiz, J.; Sastre, J.L.; Delgado, R.

1998-01-01

To prepare a common acting plan with respect to Severe Accident Management, in 1994 was founded the severe accident management ''ad-hoc'' working group from the Spanish Westinghouse PWR Nuclear Power Plant Owners Group. In this group actively collaborated the Jose Cabrera NPP Training Centre and the Department of Nuclear Engineering of UNION FENOSA. From this moment, Jose Cabrera NPP began the planning of its specific Severe Accident Management Program, which main point are Severe Accident Management Guidelines (SAMG). To elaborate this guidelines, the Spanish translation of Westinghouse Owners Group (WOG) Severe Accident Management Guidelines were considered the reference documents. The implementation of this Guidelines to Jose Cabrera NPP started on January 1997. Once the specific guidelines have been implemented to the plant, training activities for the personnel involved in severe accident issues will be developed. To prepare the training exercises MAAP4 code will be used, and with this intention, a specific Jose Cabrera NPP MAAP-GRAAPH screen has been developed. Furthermore, a wide selection of MAAP input files for the simulation of different scenarios and accidental events is available. (Author)

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).

Radioactive fallout from the Chernobyl nuclear reactor accident

International Nuclear Information System (INIS)

Beiriger, J.M.; Failor, R.A.; Marsh, K.V.; Shaw, G.E.

1987-08-01

This report describes the detection of fallout in the United States from the Chernobyl nuclear reactor accident. As part of its environmental surveillance program, Lawrence Livermore National Laboratory maintained detectors for gamma-emitting radionuclides. Following the reactor accident, additional air filters were set out. Several uncommon isotopes were detected at the time the plume passed into the US

Identification and evaluation of PWR in-vessel severe accident management strategies

International Nuclear Information System (INIS)

Dukelow, J.S.; Harrison, D.G.; Morgenstern, M.

1992-03-01

This reports documents work performed the NRC/RES Accident Management Guidance Program to evaluate possible strategies for mitigating the consequences of PWR severe accidents. The selection and evaluation of strategies was limited to the in-vessel phase of the severe accident, i.e., after the initiation of core degradation and prior to RPV failure. A parallel project at BNL has been considering strategies applicable to the ex-vessel phase of PWR severe accidents

Reliability of double-wall-tube steam generator for FBR considering water leak accident frequency

International Nuclear Information System (INIS)

Ueda, Nobuyuki; Kinoshita, Izumi; Nishi, Yoshihisa

2000-01-01

For early realization, a fast breeder reactor (FBR) is required to reduce construction cost. A reactor concept in which the intermediate heat transport system is eliminated by introducing a double-wall-tube steam generator is one convincing approach. The reliability of the double-wall-tube SG in a water leak accident (sodium-water reaction accident) due to tube failure is strongly related to the mitigating system design. The safety design of the double-wall-tube SG approach is investigated to limit the accident occurrence below 10 -7 (1/ry. A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to effectively prevent adhesion of the double-wall-tube. The reliability of the tube-to-tube plate was evaluated at 10 -10 (l/hr) for an inner tube and 10 -9 (l/hr) for an outer with reference to the failure experience of previous SGs. The failure must be detected within 30 to 60 minutes. (author)

Design features which mitigate severe accident challenges in the GE ABWR and SBWR

International Nuclear Information System (INIS)

Buchholz, Carol E.

2004-01-01

A reduction of the requirements for the emergency planning zone (EPZ) is a goal of advanced light water reactors. The technical basis for reducing the EPZ requirements is based on a very low frequency of a severe accident and high confidence that the offsite dose would be low even if a severe accident was to occur. Design features have been included in both the ABWR and SBWR to ensure that both of these goals are achieved. Probabilistic Risk Assessments (PRAs) have been performed for both plants. The PRAs indicate a core damage frequency on the order of IE-7 for both plants. The PRAs also show that the containments will not fail even if a severe accident should occur. The potential offsite is extremely low. (author)

Regulatory analyses for severe accident issues: an example

International Nuclear Information System (INIS)

Burke, R.P.; Strip, D.R.; Aldrich, D.C.

1984-09-01

This report presents the results of an effort to develop a regulatory analysis methodology and presentation format to provide information for regulatory decision-making related to severe accident issues. Insights and conclusions gained from an example analysis are presented. The example analysis draws upon information generated in several previous and current NRC research programs (the Severe Accident Risk Reduction Program (SARRP), Accident Sequence Evaluation Program (ASEP), Value-Impact Handbook, Economic Risk Analyses, and studies of Vented Containment Systems and Alternative Decay Heat Removal Systems) to perform preliminary value-impact analyses on the installation of either a vented containment system or an alternative decay heat removal system at the Peach Bottom No. 2 plant. The results presented in this report are first-cut estimates, and are presented only for illustrative purposes in the context of this document. This study should serve to focus discussion on issues relating to the type of information, the appropriate level of detail, and the presentation format which would make a regulatory analysis most useful in the decisionmaking process

Formulating the Canadian regulatory position on severe accidents

International Nuclear Information System (INIS)

Viktorov, Alex

2006-01-01

In response to the increasing potential of new nuclear build in Canada, and as part of documentation harmonization effort, CNSC staff has initiated development of requirements for design of nuclear power plants. These requirements build both on the IAEA standards, most notably, NS-R-1, and the Canadian practices and experience. The three safety objectives, formulated by the IAEA, are adopted, and Safety Goals are proposed consistent with the international trend. This Canadian standard will require, for the first time, explicit consideration of severe accidents in design and safety assessments. Specific requirements are formulated for several plant systems that assure an effective fourth level of defence in depth. Available results from probabilistic safety assessments indicate that the risks posed by severe accidents are acceptably low. Nevertheless, such risks are not negligible. CNSC staff considers that severe accident management (SAM) represents the most practical way to achieve risk reduction with a moderate effort. Ultimately, SAM actions are aimed at bringing the reactor, and the plant in general, into a controlled and stable state. For the operating reactors, SAM provides an additional defense barrier against the consequences of those accidents that fall beyond the scope of events considered in the reactor design basis. The establishment of a SAM program ensures availability of the information, procedures, and resources necessary to take full advantage of existing plant capabilities to arrest core degradation, and prevent or mitigate large releases of radioactive material. To the extent practicable, a SAM program builds on the existing emergency operating procedures and makes use of the plant design capabilities. On this basis, the CNSC requested nuclear power reactor licensees to develop and implement SAM at all operating reactors. To be able to demonstrate compliance with requirements for plant design and severe accident management, it is necessary to

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.

Proposal of the concept of selection of accidents that release large amounts of radioactive substances in the high temperature engineering test reactor

International Nuclear Information System (INIS)

Ono, Masato; Honda, Yuki; Takada, Shoji; Sawa, Kazuhiro

2015-01-01

In Position, construction and equipment of testing and research reactor to be subjected to the use standards for rules Article 53 (prevention of expansion of the accident to release a large amount of radioactive material) generation the frequency is a lower accident than design basis accident, when what is likely to release a large amount of radioactive material or radiation from the facility has occurred, and take the necessary measures in order to prevent the spread of the accident. There is provided a lower accident than frequency design basis accidents, for those that may release a large amount of radioactive material or radiation. (author)

NPP Krsko Severe Accident Management Guidelines Upgrade

International Nuclear Information System (INIS)

Mihalina, Mario; Spalj, Srdjan; Glaser, Bruno; Jalovec, Robi; Jankovic, Gordan

2014-01-01

Nuclear Power Plant Krsko (NEK) has decided to take steps for upgrade of safety measures to prevent severe accidents, and to improve the means to successfully mitigate their consequences. The content of the program for the NEK Safety Upgrade is consistent with the nuclear industry response to Fukushima accident, which revealed many new insights into severe accidents. Therefore, new strategies and usage of new systems and components should be integrated into current NEK Severe Accident Management Guidelines (SAMG's). SAMG's are developed to arrest the progression of a core damage accident and to limit the extent of resulting releases of fission products. NEK new SAMG's revision major changes are made due to: replacement of Electrical Recombiners by Passive Autocatalytic Recombiners (PARs) and the installation of Passive Containment Filtered Vent System (PCFV); to handle a fuel damage situation in Spent Fuel Pool (SFP) and to assess risk of core damage situation during shutdown operation. (authors)

Transportation accidents/incidents involving radioactive materials (1971--1991)

International Nuclear Information System (INIS)

Cashwell, C.E.; McClure, J.D.

1992-01-01

The Radioactive Materials Incident Report (RMIR) database contains information on transportation-related accidents and incidents involving radioactive materials that have occurred in the United States. The RMIR was developed at Sandia National Laboratories (SNL) to support its research and development program efforts for the US Department of Energy (DOE). This paper will address the following topics: background information on the regulations and process for reporting a hazardous materials transportation incident, overview data of radioactive materials transportation accidents and incidents, and additional information and summary data on how packagings have performed in accident conditions

Severe Accident Management System On-line Network SAMSON

International Nuclear Information System (INIS)

Silverman, Eugene B.

2004-01-01

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

The incidence of multiple fatality accidents in the U.K

International Nuclear Information System (INIS)

Griffiths, R.F.; Fryer, L.S.

1978-03-01

The report presents a compilation of available data showing the frequency of occurrence of accidents resulting in multiple fatalities in the United Kingdom (or Great Britain, as the data permits). Frequency vs. number-of-fatalities graphs are plotted for the following five classes of fatalities:- persons killed in fires in the UK; crew members of UK registered ships; railway passengers in Great Britain; passengers of aircraft on the British register; occupants of Public Service Vehicles in Great Britain. The information needed to compile a more comprehensive set of graphs was found to be unobtainable. The report concludes with the suggestion that procedures for reporting and recording details of fatal accidents should be reviewed. (author)

Description of the blowdown test facility COG program on in-reactor fission product release, transport, and deposition under severe accident conditions

International Nuclear Information System (INIS)

Fehrenbach, P.J.; Wood, J.C.

1987-06-01

Loss-of-coolant accidents with additional impairment of emergency cooling would probably result in high fuel temperatures leading to severe fuel damage (SFD) and significant fission product activity would then be transported along the PHTS to the break where a fraction of it would be released and transport under such conditions, there are many interacting and sometimes competing phenomena to consider. Laboratory simulations are being used to provide data on these individual phenomena, such as UO 2 oxidation and Zr-UO 2 interaction, from which mathematical models can be constructed. These are then combined into computer codes to include the interaction effects and assess the overall releases. In addition, in-reactor tests are the only source of data on release and transport of short-lived fission product nuclides, which are important in the consequence analysis of CANDU reactor accidents. Post-test decontamination of an in-reactor test facility also provides a unique opportunity to demonstrate techniques and obtain decontamination data relevant to post-accident rehabilitation of CANDU power reactors. Specialized facilities are required for in-reactor testing because of the extensive release of radioactive fission products and the high temperatures involved (up to 2500 degrees Celsius). To meet this need for the Canadian program, the Blowdown Test Facility (BTF) has been built in the NRU reactor at Chalk River. Between completion of construction in mid-1987 and the first Zircaloy-sheathed fuel test in fiscal year 1987/88, several commissioning tests are being performed. Similarly, extensive development work has been completed to permit application of instrumentation to irradiated fuel elements, and in support of post-test fuel assembly examination. A program of decontamination studies has also been developed to generate information relevant to post-accident decontamination of power reactors. The BTF shared cost test program funded by the COG High Temperature

Comparison of skin sensory thresholds using pre-programmed or single-frequency transcutaneous electrical nerve stimulation.

Science.gov (United States)

Kang, Jong Ho

2015-12-01

[Purpose] The purpose of the present study was to compare the sensory thresholds of healthy subjects using pre-programmed or single-frequency transcutaneous electrical nerve stimulation. [Subjects] Ninety healthy adult subjects were randomly assigned to pre-programmed or single-frequency stimulation groups, each consisting of 45 participants. [Methods] Sensory thresholds were measured in the participants' forearms using von Frey filaments before and after pre-programmed or single-frequency transcutaneous electrical nerve stimulation, and the result in values were analyzed. [Results] Significant increases in sensory threshold after stimulation were observed in both groups. However, there were no significant differences between the two groups in sensory thresholds after stimulation or in the magnitude of threshold increases following stimulation. [Conclusion] Our results show that there are no differences between sensory threshold increases induced by pre-programmed and single-frequency transcutaneous electrical nerve stimulation.

Instrumentation Performance during the TMI-2 Accident

International Nuclear Information System (INIS)

Rempe, Joy L.; Knudson, Darrell L.

2013-06-01

The accident at the Three Mile Island Unit 2 (TMI- 2) reactor provided a unique opportunity to evaluate sensors exposed to severe accident conditions. Conditions associated with the release of coolant and the hydrogen burn that occurred during this accident exposed instrumentation to harsh conditions, including direct radiation, radioactive contamination, and high humidity with elevated temperatures and pressures. As part of a program initiated by the Department of Energy Office of Nuclear Energy (DOE-NE), a review was completed to gain insights from prior TMI-2 sensor survivability and data qualification efforts. This new effort focused upon a set of sensors that provided critical data to TMI-2 operators for assessing the condition of the plant and the effects of mitigating actions taken by these operators. In addition, the effort considered sensors providing data required for subsequent accident simulations. Over 100 references related to instrumentation performance and post-accident evaluations of TMI-2 sensors and measurements were reviewed. Insights gained from this review are summarized within this paper. As noted within this paper, several techniques were invoked in the TMI-2 post-accident program to evaluate sensor survivability status and data qualification, including comparisons with data from other sensors, analytical calculations, laboratory testing, and comparisons with sensors subjected to similar conditions in large-scale integral tests and with sensors that were similar in design but more easily removed from the TMI-2 plant for evaluations. Conclusions from this review provide important insights related to sensor survivability and enhancement options for improving sensor performance. In addition, this paper provides recommendations related to sensor survivability and the data evaluation process that could be implemented in upcoming Fukushima Daiichi recovery efforts. (authors)

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

Accidents in radiotherapy: Lack of quality assurance?

International Nuclear Information System (INIS)

Novotny, J.

1997-01-01

About 150 radiological accidents, involving more than 3000 patients with adverse effects, 15 patient's fatalities and about 5000 staff and public exposures have been collected and analysed. Out of 67 analysed accidents in external beam therapy 22% has been caused by wrong calculation of the exposure time or monitor units, 13% by inadequate review of patient's chart, 12% by mistakes in the anatomical area to be treated. The remaining 35% can be attributed to 17 different causes. The most common mistakes in brachytherapy were wrong activities of sources used for treatment (20%), inadequate procedures for placement of sources applicators (14%), mistakes in calculating the treatment time (12%), etc. The direct and contributing causes of radiological accidents have been deduced from each event, when it was possible and categorized into 9 categories: mistakes in procedures (30%), professional mistakes (17%), communication mistakes (15%), lack of training (8.5%), interpretation mistakes (7%), lack of supervision (6%), mistakes in judgement (6%), hardware failures (5%), software and other mistakes (5.5%). Three types of direct and contributing causes responsible for almost 62% of all accidents are directly connected to the quality assurance of treatment. The lessons learnt from the accidents are related to frequencies of direct and contributing factors and show that most of the accident are caused by lack, non-application of quality assurance (QA) procedures or by underestimating of QA procedures. The international system for collection of accidents and dissemination of lessons learnt from the different accidents, proposed by IAEA, can contribute to better practice in many radiotherapy departments. Most of the accidents could have been avoided, had a comprehensive QA programme been established and properly applied in all radiotherapy departments, whatever the size. (author)

Frequency modulation television analysis: Threshold impulse analysis. [with computer program

Science.gov (United States)

Hodge, W. H.

1973-01-01

A computer program is developed to calculate the FM threshold impulse rates as a function of the carrier-to-noise ratio for a specified FM system. The system parameters and a vector of 1024 integers, representing the probability density of the modulating voltage, are required as input parameters. The computer program is utilized to calculate threshold impulse rates for twenty-four sets of measured probability data supplied by NASA and for sinusoidal and Gaussian modulating waveforms. As a result of the analysis several conclusions are drawn: (1) The use of preemphasis in an FM television system improves the threshold by reducing the impulse rate. (2) Sinusoidal modulation produces a total impulse rate which is a practical upper bound for the impulse rates of TV signals providing the same peak deviations. (3) As the moment of the FM spectrum about the center frequency of the predetection filter increases, the impulse rate tends to increase. (4) A spectrum having an expected frequency above (below) the center frequency of the predetection filter produces a higher negative (positive) than positive (negative) impulse rate.

Evaluation of severe accident safety system value based on averting financial risks

International Nuclear Information System (INIS)

Hatch, S.W.; Benjamin, A.S.; Bennett, P.R.

1983-01-01

The Severe Accident Risk Reduction Program is being performed to benchmark the risks from nuclear power plants and to assess the benefits and impacts of a set of severe accident safety features. This paper describes the program in general and presents some preliminary results. These results include estimates of the financial risks associated with the operation of six reference plants and the value of severe accident prevention and mitigation safety systems in averting these risks. The results represent initial calculations and will be iterated before being used to support NRC decisions

System 80+ design features for severe accident prevention and mitigation

International Nuclear Information System (INIS)

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

1993-01-01

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

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

International Nuclear Information System (INIS)

Hsii, Y.H.

1976-06-01

The CONTEMPT computer program predicts the pressure-temperature response of a single-volume reactor building to a loss-of-coolant accident. The report describes the analytical model used for the program. CONTEMPT assumes that the loss-of-coolant accident can be separated into two phases; the primary system blowdown and reactor building pressurization. The results of the blowdown analysis serve as the boundary conditions and are input to the CONTEMPT program. Thus, the containment model is only concerned with the pressure and temperature in the reactor building and the temperature distribution through the reactor building structures. The user is required to input the description of the discharge of coolant, the boiling of residual water by reactor decay heat, the superheating of steam passing through the core, and metal-water reactions. The reactor building is separated into liquid and vapor regions. Each region is in thermal equilibrium itself, but the two may not be in thermal equilibrium; the liquid and gaseous regions may have different temperatures. The reactor building is represented as consisting of several heat-conducting structures whose thermal behavior can be described by the one-dimensional multi-region heat conduction equation. The program also calculates building leakage and the effects of engineered safety features such as reactor building sprays, decay heat coolers, sump coolers, etc

Hazards to nuclear plants from surface traffic accidents

International Nuclear Information System (INIS)

Hornyik, K.

1975-01-01

Analytic models have been developed for evaluating hazards to nuclear plants from hazardous-materials accidents in the vicinity of the plant. In particular, these models permit the evaluation of hazards from such accidents occurring on surface traffic routes near the plant. The analysis uses statistical information on accident rates, traffic frequency, and cargo-size distribution along with parameters describing properties of the hazardous cargo, plant design, and atmospheric conditions, to arrive at a conservative estimate of the annual probability of a catastrophic event. Two of the major effects associated with hazardous-materials accidents, explosion and release of toxic vapors, are treated by a common formalism which can be readily applied to any given case by means of a graphic procedure. As an example, for a typical case it is found that railroad shipments of chlorine in 55-ton tank cars constitute a greater hazard to a nearby nuclear plant than equally frequent rail shipments of explosives in amounts of 10 tons. 11 references. (U.S.)

Swedish REGULATORY APPROACH TO SAFETY Assessment AND SEVERE ACCIDENT MANAGEMENT

International Nuclear Information System (INIS)

Frid, W.; Sandervaag, O.

1997-01-01

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

[Self-reported substance abuse related emergencies: frequency and nature].

Science.gov (United States)

Schäfer, G; Smoltczyk, H; Dengler, W; Buchkremer, G

2000-04-01

The aim of this study was to estimate the frequency and nature of self-reported and drug-related emergencies. 47 patients of a ward for opiate detoxification were interviewed about their experiences with drug-related emergencies. Typical categories had to be found like overdoses, seizures, accidents and suicide attempts respectively. 68% had own experience with drug-related emergency. A majority suffered opiate overdose with different extensions as unconsciousness or breath-depression. Alcohol and polydrug use was associated with overdose. Drug-related accidents were only reported by men. Half the number of drug-related emergencies were treated in hospital. Most emergencies occurred alone either in a home environment or outside. Harm reduction interventions like observed user rooms should be established. Furthermore other strategies to reduce the number of emergencies as sharing naloxon or resuscitation programs in wards for detoxification could also be an effective method to prevent near fatal or fatal overdoses in dependent subjects.

Thermal hydraulic behavior of a PWR under beyond-design-basis accident conditions: Conclusions from an experimental program in a 4-loop test facility (PKL)

International Nuclear Information System (INIS)

Umminger, K.J.; Kastner, W.; Mandl, R.M.; Weber, P.

1993-01-01

Within the scope of German reactor safety research, extensive experiments covering the behavior of nuclear power plants under accident conditions have been carried out in the PKL test facility which simulates a 4-loop, 1,300 MWe KWU-designed PWR. While the investigations dealing with design-basis accidents and with the efficiency of the emergency core cooling systems have been largely completed, the main interest nowadays concentrates on the investigation of beyond-design-basis accidents to demonstrate the safety margins of nuclear power plants and to investigate the contribution of the built-in safety features for a further reduction of the residual risk. The thermal hydraulic behavior of a PWR under these extreme accident conditions was experimentally investigated within the PKL III B test program. This paper presents the fundamental findings with some of the most important results being discussed in detail. Future plans are also outlined

URBAN TRAFFIC ACCIDENT ANALYSIS BY USING GEOGRAPHIC INFORMATION SYSTEM

Directory of Open Access Journals (Sweden)

Meltem SAPLIOĞLU

2006-03-01

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

Accident scenarios triggered by lightning strike on atmospheric storage tanks

International Nuclear Information System (INIS)

Necci, Amos; Argenti, Francesca; Landucci, Gabriele; Cozzani, Valerio

2014-01-01

Severe Natech accidents may be triggered by lightning strike affecting storage tanks containing relevant inventories of hazardous materials. The present study focused on the identification of event sequences and accident scenarios following lightning impact on atmospheric tanks. Reference event trees, validated using past accident analysis, are provided to describe the specific accident chains identified, accounting for reference protection and mitigation safety barriers usually adopted in current industrial practice. An overall methodology was outlined to allow the calculation of the expected frequencies of final scenarios following lightning impact on atmospheric storage tanks, taking into account the expected performance of available safety barriers. The methodology was applied to a case study in order to better understand the data that may be obtained and their importance in the framework of quantitative risk assessment (QRA) and of the risk management of industrial facilities with respect to external hazards due to natural events. - Highlights: • Event sequences following lightning impact on atmospheric tanks were identified. • Reference event trees including standard safety barriers were obtained. • Safety barriers applied in industrial practice were assessed to quantify event trees. • Frequencies of final scenarios following lightning impact on tanks were calculated. • Natech scenarios caused by lightning have an important influence on risk profiles

Statistical inference of the nuclear accidents occurrence number for the next decade

International Nuclear Information System (INIS)

Felizia, E.R.

1987-01-01

This paper aims to give a response using the classical statistical and bayesian inference techniques regarding the common characteristic in the Harrisburg and Chernobyl nuclear accidents: in both reactors, core fusion occurred. In relation to the last mentioned techniques, the most recent developments were applied, based on the decision theory of uncertainty; among others, the principle of maximum entropy. Besides, as a preliminar information on the accidents occurrence frequency with core fusion, the German risk analysis results were used. The estimations predicted for the next decade an average between one or two accidents with core fusion and low possibilities for the 'no accident' event in the same period. (Author)

A brief review of the accident at Three Mile Island

International Nuclear Information System (INIS)

Corey, G.R.

1979-01-01

A question-and-answer format is used to discuss the Three Mile Reactor accident and the lessons learned. The aspects touched upon include the hydrogen bubble, the radiation levels the public was exposed to, and the consequences of the accident to the nuclear power program

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

Containment performance improvement program

International Nuclear Information System (INIS)

Beckner, W.; Mitchell, J.; Soffer, L.; Chow, E.; Lane, J.; Ridgely, J.

1990-01-01

The Containment Performance Improvement (CPI) program has been one of the main elements in the US Nuclear Regulatory Commission's (NRC's) integrated approach to closure of severe accident issues for US nuclear power plants. During the course of the program, results from various probabilistic risk assessment (PRA) studies and from severe accident research programs for the five US containment types have been examined to identify significant containment challenges and to evaluate potential improvements. The five containment types considered are: the boiling water reactor (BMR) Mark I containment, the BWR Mark II containment, the BWR Mark III containment, the pressurized water reactor (PWR) ice condenser containment, and the PWR dry containments (including both subatmospheric and large subtypes). The focus of the CPI program has been containment performance and accident mitigation, however, insights are also being obtained in the areas of accident prevention and accident management

Unrecorded Accidents Detection on Highways Based on Temporal Data Mining

Directory of Open Access Journals (Sweden)

Shi An

2014-01-01

Full Text Available Automatic traffic accident detection, especially not recorded by traffic police, is crucial to accident black spots identification and traffic safety. A new method of detecting traffic accidents is proposed based on temporal data mining, which can identify the unknown and unrecorded accidents by traffic police. Time series model was constructed using ternary numbers to reflect the state of traffic flow based on cell transmission model. In order to deal with the aftereffects of linear drift between time series and to reduce the computational cost, discrete Fourier transform was implemented to turn time series from time domain to frequency domain. The pattern of the time series when an accident happened could be recognized using the historical crash data. Then taking Euclidean distance as the similarity evaluation function, similarity data mining of the transformed time series was carried out. If the result was less than the given threshold, the two time series were similar and an accident happened probably. A numerical example was carried out and the results verified the effectiveness of the proposed method.

Radiation accidents: occurrence, types, consequences, medical management, and the lessons to be learned

International Nuclear Information System (INIS)

Turai, I.; Veress, K.

2001-01-01

The paper reviews the frequency, causes and occurrence of radiation accidents with some significant exposure to human. More detailed information is provided in tabulated form on the health consequences of those twenty severe radiation accidents that occurred in 1986-2000, world-wide. Reference is given to the very low cumulative incidence of significant radiation accidents, as during the last 57 years there were, in average, seven registered accidents annually in all countries of the world. Thus, the chance for most of the physicians to meet a patient with symptoms of acute radiation injury during their professional career is very low

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)

[Factors associated with road accidents in the city of Lubumbashi].

Science.gov (United States)

Kandolo, Simon Ilunga; Matungulu, Charles Matungulu; Mukanya, Pascal Kimba; Umba, Irène Kabange; Kabamba, Julie Ndayi; Numbi, Oscar Luboya; Ilunga, Benjamin Kabyla; Kaj, Françoise Malonga; Lubaba, Célestin Banza; Ngongo, Ghislain Mashini

2014-01-01

We conducted a retrospective non-comparative cross-sectional study in Lubumbashi on factors associated with road accidents. To contribute to the reduction of morbidity and mortality related to road accidents based on a better understanding of this phenomenon in the city of Lubumbashi, to assess the frequency and the growth of road accidents in the city of Lubumbashi and to determine the factors predisposing to these accidents. This non-controlled retrospective cross-sectional descriptive study covered the period from 1st January to 31st December 2008, and included all cases of vehicles registered by the police department after an accident. A total of 1,915 road traffic were reported, mostly caused by public transport vehicles (59%) driven by young adult males (89% of cases) with a mean age of 28 years. The main cause was speeding combined with other factors (34% of cases). The majority of these accidents occurred in the city of Lubumbashi (52.9% of cases) between 6:00 a.m. and 6:00 p.m. (74% of cases). Transit vehicles were exposed to a greater risk of road accidents (59%). There is an urgent need for accurate information on the rules of the road for drivers and road users to reduce the mortality and morbidity due to road accidents.

PREVENTION OF OCCUPATIONAL ACCIDENTS

Directory of Open Access Journals (Sweden)

Jovica Jovanovic

2004-01-01

Full Text Available Medical services, physicians and nurses play an essential role in the plant safety program through primary treatment of injured workers and by helping to identify workplace hazards. The physician and nurse should participate in the worksite investigations to identify specific hazard or stresses potentially causing the occupational accidents and injuries and in planning the subsequent hazard control program. Physicians and nurses must work closely and cooperatively with supervisors to ensure the prompt reporting and treatment of all work related health and safety problems. Occupational accidents, work related injuries and fatalities result from multiple causes, affect different segments of the working population, and occur in a myriad of occupations and industrial settings. Multiple factors and risks contribute to traumatic injuries, such as hazardous exposures, workplace and process design, work organization and environment, economics, and other social factors. With such a diversity of theories, it will not be difficult to understand that there does not exist one single theory that is considered right or correct and is universally accepted. These theories are nonetheless necessary, but not sufficient, for developing a frame of reference for understanding accident occurrences. Prevention strategies are also varied, and multiple strategies may be applicable to many settings, including engineering controls, protective equipment and technologies, management commitment to and investment in safety, regulatory controls, and education and training. Research needs are thus broad, and the development and application of interventions involve many disciplines and organizations.

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...

Agricultural countermeasure program - AGRICP: food and dose module in ARGOS- accident reporting and Guidance Operational System

International Nuclear Information System (INIS)

Calábria, Jaqueline A.A.; Morais, Gustavo F.

2017-01-01

Nuclear or radiological emergencies can affect food, feed and commodities grown. The regulatory bodies has a role in the post-accident phase instructing the population regarding the consumption of agricultural products, monitoring and recovering the contaminated areas and disposing the generated waste. To deal with nuclear/radiological emergencies, in the end of 2007, Brazil took part of the ARGOS consortium. ARGOS is a software used for support the Preparedness and Response of a nuclear emergency. Specifically for use during the recovery phase, ARGOS has a module called AgriCP (Agricultural Countermeasure Program). This functionality was add to the version 9.0 of ARGOS, in 2012, replacing FMD (Food and Dose Module) model. AgriCP can be very useful in the post-accident phasing, helping to planning the actions that must be taken, saving human and budged resources. However, most of the parameters used by default for the model are specific for Central Europe and must be adapted to the Brazilian characteristics. In this paper the basic functionalities of AgriCP are presented and a general view of the issues to be addressed while implementing AgriCP for the Brazilian case is given. Besides the lack of specific parameters for the Brazilian reality, the definition of the area to be considering for intervention in an accident, taking into account the very complex meteorological characteristic of the Brazilian NPPs (nuclear power plants) site, are some of the matters of concern. (author)

Agricultural countermeasure program - AGRICP: food and dose module in ARGOS- accident reporting and Guidance Operational System

Energy Technology Data Exchange (ETDEWEB)

Calábria, Jaqueline A.A.; Morais, Gustavo F., E-mail: jaqueline.calabria@cnen.gov.br, E-mail: gustavo.morais@cnen.gov.br [Comissao Nacional de Energia Nuclear (CNEN), Rio de Janeiro, RJ (Brazil)

2017-11-01

Nuclear or radiological emergencies can affect food, feed and commodities grown. The regulatory bodies has a role in the post-accident phase instructing the population regarding the consumption of agricultural products, monitoring and recovering the contaminated areas and disposing the generated waste. To deal with nuclear/radiological emergencies, in the end of 2007, Brazil took part of the ARGOS consortium. ARGOS is a software used for support the Preparedness and Response of a nuclear emergency. Specifically for use during the recovery phase, ARGOS has a module called AgriCP (Agricultural Countermeasure Program). This functionality was add to the version 9.0 of ARGOS, in 2012, replacing FMD (Food and Dose Module) model. AgriCP can be very useful in the post-accident phasing, helping to planning the actions that must be taken, saving human and budged resources. However, most of the parameters used by default for the model are specific for Central Europe and must be adapted to the Brazilian characteristics. In this paper the basic functionalities of AgriCP are presented and a general view of the issues to be addressed while implementing AgriCP for the Brazilian case is given. Besides the lack of specific parameters for the Brazilian reality, the definition of the area to be considering for intervention in an accident, taking into account the very complex meteorological characteristic of the Brazilian NPPs (nuclear power plants) site, are some of the matters of concern. (author)

Global ship accidents and ocean swell-related sea states

Directory of Open Access Journals (Sweden)

Z. Zhang

2017-11-01

Full Text Available With the increased frequency of shipping activities, navigation safety has become a major concern, especially when economic losses, human casualties and environmental issues are considered. As a contributing factor, the sea state plays a significant role in shipping safety. However, the types of dangerous sea states that trigger serious shipping accidents are not well understood. To address this issue, we analyzed the sea state characteristics during ship accidents that occurred in poor weather or heavy seas based on a 10-year ship accident dataset. Sea state parameters of a numerical wave model, i.e., significant wave height, mean wave period and mean wave direction, were analyzed for the selected ship accident cases. The results indicated that complex sea states with the co-occurrence of wind sea and swell conditions represent threats to sailing vessels, especially when these conditions include similar wave periods and oblique wave directions.

Global ship accidents and ocean swell-related sea states

Science.gov (United States)

Zhang, Zhiwei; Li, Xiao-Ming

2017-11-01

With the increased frequency of shipping activities, navigation safety has become a major concern, especially when economic losses, human casualties and environmental issues are considered. As a contributing factor, the sea state plays a significant role in shipping safety. However, the types of dangerous sea states that trigger serious shipping accidents are not well understood. To address this issue, we analyzed the sea state characteristics during ship accidents that occurred in poor weather or heavy seas based on a 10-year ship accident dataset. Sea state parameters of a numerical wave model, i.e., significant wave height, mean wave period and mean wave direction, were analyzed for the selected ship accident cases. The results indicated that complex sea states with the co-occurrence of wind sea and swell conditions represent threats to sailing vessels, especially when these conditions include similar wave periods and oblique wave directions.

Development of an accident management expert system for containment assessment

International Nuclear Information System (INIS)

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

1987-01-01

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

Analysis of pressurized water reactor accidents in reactivity disturbances. II

International Nuclear Information System (INIS)

Tinka, I.

1978-01-01

The logic structure of program FATRAP is described. The time course of reactivity temporal and spatial distributions of neutron flux density and power, characteristic temperatures of the individual reactor zones and the heat flux density from cladding to the coolant can be obtained as the main results. The basic program funcitons were tested for a point and a one-dimensional model. In the basic test the absorption rod was removed uncontrollably at a preset speed for 0.5 s with the reactivity feedback operative. A second test simulated the action of the accident protection system with a delay of 0.1 s started when the 7500 MW power had been obtained. The last test consisted in simulating a start-up accident with an initial power of 2.25 MW. For the said chosen accident models reactivity feedback is responsible for the formation of the appropriate power peak while the accident protection attendance alone can considerably reduce temperatures during the process. (J.F.)

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.

PSA modeling of long-term accident sequences

International Nuclear Information System (INIS)

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

2014-01-01

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

DOZIM - evaluation dose code for nuclear accident

International Nuclear Information System (INIS)

Oprea, I.; Musat, D.; Ionita, I.

2008-01-01

During a nuclear accident an environmentally significant fission products release can happen. In that case it is not possible to determine precisely the air fission products concentration and, consequently, the estimated doses will be affected by certain errors. The stringent requirement to cope with a nuclear accident, even minor, imposes creation of a computation method for emergency dosimetric evaluations needed to compare the measurement data to certain reference levels, previously established. These comparisons will allow a qualified option regarding the necessary actions to diminish the accident effects. DOZIM code estimates the soil contamination and the irradiation doses produced either by radioactive plume or by soil contamination. Irradiations either on whole body or on certain organs, as well as internal contamination doses produced by isotope inhalation during radioactive plume crossing are taken into account. The calculus does not consider neither the internal contamination produced by contaminated food consumption, or that produced by radioactive deposits resuspension. The code is recommended for dose computation on the wind direction, at distances from 10 2 to 2 x 10 4 m. The DOZIM code was utilized for three different cases: - In air TRIGA-SSR fuel bundle destruction with different input data for fission products fractions released into the environment; - Chernobyl-like accident doses estimation; - Intervention areas determination for a hypothetical severe accident at Cernavoda Nuclear Power Plant. For the first case input data and results (for a 60 m emission height without iodine retention on active coal filters) are presented. To summarize, the DOZIM code conception allows the dose estimation for any nuclear accident. Fission products inventory, released fractions, emission conditions, atmospherical and geographical parameters are the input data. Dosimetric factors are included in the program. The program is in FORTRAN IV language and was run on

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

International Nuclear Information System (INIS)

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

2003-01-01

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

Risk assessment of DOE defense program packages in a beyond 10 CFR 71.73 transportation accident environment

International Nuclear Information System (INIS)

Sandquist, G.M.; Bennion, J.S.; Moore, J.E.

1992-01-01

A comprehensive program is being conducted by the DOE to determine the risks related to the domestic transportation of radioactive and hazardous materials associated with nuclear weapons. The program is designed to identify, quantify and manage potential risks to public health and safety including potential radiological and toxicological health consequences which may exceed the 10 CFR 71.73 transportation accident environment A major objective of this program being performed by the Lawrence Livermore National Laboratory (LLNL) and the University of Utah is to provide the DOE with the methodology and bases for evaluating highway transportation activities by DOE contractors. This paper describes the approach and the HITRA model which is based upon probabilistic risk assessment (PRA) methodology and route specific data associated with the proposed transportation activity. The model is capable of providing detailed, location and time specific data for assessing projected risks to public health and safety from DOE defense program materials shipments

Aspects of accident management in Cernavoda NPP

International Nuclear Information System (INIS)

Dascalu, N.

1999-01-01

As a general conclusion, the accident management system as implemented at Cerna voda NPP is expected to be appropriate for handling a severe accident, should it occur, in such a way that the environmental radiological consequences would be insignificant and radiation exposure of the personnel be within recommendations. It is recognized, however, that continued development and verification of the system as well as effective personnel training programs are essential to maintain the safety level achieved. (author)

Results and tasks of the implementation of federal target programs aimed at overcoming the consequences of radiation accidents and catastrophes in the Russian Federation

International Nuclear Information System (INIS)

Gerasimova, N.V.

2002-01-01

Major results are presented on the implementation of federal target programs on overcoming the consequences of the accident at the Chernobyl nuclear power plant, radiation accidents and incidents at the 'Mayak' Industrial Association, nuclear tests at the Semipalatinsk test site in the period of 1992-2000. The status of the standards and legislation regulating the activities aimed at population protection and rehabilitation of territories is analyzed. The current state of the problem is evaluated. The proposals are laid down for major directions of the state policy of the Russian Federation in overcoming the consequences of radiation accidents for the period until 2010, and the outlook for the efforts in the above domain and the above period is given. About 130 thousand square kilometers of the territories of 20 Russian Federation subjects with a population of around 4 million people were affected by accidents at nuclear fuel cycle sites/facilities, and nuclear and hydrogen weapons tests. The accidents entailed a host of grave radioecological, medical, demographic, and socio-economic consequences, exerted a significant unfavorable impact upon the socio-economic development of the affected territories. (author)

Full-length fuel rod behavior under severe accident conditions

International Nuclear Information System (INIS)

Lombardo, N.J.; Lanning, D.D.; Panisko, F.E.

1992-12-01

This document presents an assessment of the severe accident phenomena observed from four Full-Length High-Temperature (FLHT) tests that were performed by the Pacific Northwest Laboratory (PNL) in the National Research Universal (NRU) reactor at Chalk River, Ontario, Canada. These tests were conducted for the US Nuclear Regulatory Commission (NRC) as part of the Severe Accident Research Program. The objectives of the test were to simulate conditions and provide information on the behavior of full-length fuel rods during hypothetical, small-break, loss-of-coolant severe accidents, in commercial light water reactors

The risk of major nuclear accident: calculation and perception of probabilities

International Nuclear Information System (INIS)

Leveque, Francois

2013-01-01

Whereas before the Fukushima accident, already eight major accidents occurred in nuclear power plants, a number which is higher than that expected by experts and rather close to that corresponding of people perception of risk, the author discusses how to understand these differences and reconcile observations, objective probability of accidents and subjective assessment of risks, why experts have been over-optimistic, whether public opinion is irrational regarding nuclear risk, and how to measure risk and its perception. Thus, he addresses and discusses the following issues: risk calculation (cost, calculated frequency of major accident, bias between the number of observed accidents and model predictions), perceived probabilities and aversion for disasters (perception biases of probability, perception biases unfavourable to nuclear), the Bayes contribution and its application (Bayes-Laplace law, statistics, choice of an a priori probability, prediction of the next event, probability of a core fusion tomorrow)

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)

Investigation Effect of Biorhythm on Work-Related Accidents in The Metal Industry (A Short Report

Directory of Open Access Journals (Sweden)

Ehsanollah Habibi

2016-07-01

Full Text Available Biorhythm is one of the newest subjects in the field of cognition of mental ergonomics which can be very effective in reduction of work-related accidents or mistakes with no apparent reason. With evaluating Biorhythm individuals can intervention action to reduce job accidents carried out. Thus, the aim of this study was to determine the relationship Biorhythm and work-related accidents in the metal industry. This research is a cross-sectional and analytical-descriptive in the metal industrial Isfahan city of 120 work-related accidents during 2015. The required information was collected from available documents in HSE unit of the company biorhythm charts were drawn based on a date of accidents and participants birthdays, using natural Biorhythm Software V3.02 Conduct. Finally، the data were analyzed using spss version 20 and descriptive statistics.This study showed that the frequency of accidents in critical days and negative section of physical cycle was more than expected. Also the frequency of accidents in critical days and negative section of emotional and intellectual cycles was less than expected. Most type of injury, including cuts to 35.8 percent and the lowest type of injury was torsion with 5 percent. Most limb injury, hands and fingers with 51.7 percent and the lowest limb injury were back at 2.5 percent. Accidents outbreak in physical cycles was 38.3 percent. These 120 accidents in additionally were causing 120 loss of working days in effect accident. Most percent of loss of working days were for 20 to 30 days with of 39.2 percent. Most percent of loss of working days were for 20 to 30 days with of 39.2 percent. Due to the physical nature of the work activities in the metal industry can be stated that the study showed that in physical work activities, frequency of accidents in critical days and negative section of physical cycle in which the person is not physically ready to do the job was more than expected. Therefore, by training

Perspectives on the economic risks of LWR accidents

International Nuclear Information System (INIS)

Ritchie, L.T.; Burke, R.P.

1986-01-01

Models which can be used for the analysis of the economic risks from events which may occur during LWR operation have been developed. The models include capabilities to estimate both onsite and offsite costs of LWR events ranging from routine plant forced outages to severe core-melt accidents resulting in large releases of radioactive material to the environment. The economic consequence models have been applied in studies of the economic risks from the operation of US LWR plants. The results of the analyses provide some important perspectives regarding the economic risks of LWR accidents. The analyses indicate that economic risks, in contrast to public health risks, are dominated by the onsite costs of relatively high-frequency forced outage events. Even for severe (e.g., core-melt) accidents, expected offsite costs are less than expected onsite costs for a typical US plant

Severe accident progression perspectives for Mark I containments based on the IPE results

International Nuclear Information System (INIS)

Lin, C.C.; Lehner, J.R.; Pratt, W.T.; Drouin, M.

1995-01-01

Based on level 2 analyses in IPE (Individual Plant Examination) submittals accident progression, perspectives were obtained for all containment types. These perspectives consisted of insights on containment failure modes, releases therein, and factors responsible for the results. To illustrate the types of perspectives acquired on severe accident progresssion, insights obtained for (BWR) Mark I containments are discussed here. Mark I containments have high strength but small volumes and rely on pressure suppression pools to condense steam released from the reactor coolant system during an accident. Accidents causing structural failure of the drywell shortly after the core debris melts through the reactor vessel were found to be dominant contributors to risk. Importance of individual containment failure mechanisms depends on plant features and in some cases on modeling assumptions; however the following mechanisms were found important: drywell shell melt-through caused by direct contact with core debris and drywell failure caused by rapid pressure/temperature pulses at time of vessel melt-through. Drywell failure caused by gradual pressure/temperature buildup due to gases and steam released during core/concrete interactions is important in some IPEs. In other IPEs vent was an important contributor. However, accidents that bypass containment (eg interfacing systems LOCA)or involve containment isolation failure were not important contributors to the CDF in any of the IPEs for Mark I plants. These accidents are also not important to risk (even though they can involve large fission product release) because their frequencies of occurrence are so much lower than frequencies of early structural failure caused by other accidents that dominate the CDF

Severe accident management guidelines tool

International Nuclear Information System (INIS)

Gutierrez Varela, Javier; Tanarro Onrubia, Augustin; Martinez Fanegas, Rafael

2014-01-01

Severe Accident is addressed by means of a great number of documents such as guidelines, calculation aids and diagnostic trees. The response methodology often requires the use of several documents at the same time while Technical Support Centre members need to assess the appropriate set of equipment within the adequate mitigation strategies. In order to facilitate the response, TECNATOM has developed SAMG TOOL, initially named GGAS TOOL, which is an easy to use computer program that clearly improves and accelerates the severe accident management. The software is designed with powerful features that allow the users to focus on the decision-making process. Consequently, SAMG TOOL significantly improves the severe accident training, ensuring a better response under a real situation. The software is already installed in several Spanish Nuclear Power Plants and trainees claim that the methodology can be followed easier with it, especially because guidelines, calculation aids, equipment information and strategies availability can be accessed immediately (authors)

Identification of the operating crew's information needs for accident management

Energy Technology Data Exchange (ETDEWEB)

Nelson, W.R.; Hanson, D.J.; Ward, L.W.; Solberg, D.E.

1988-01-01

While it would be very difficult to predetermine all of the actions required to mitigate the consequences of every potential severe accident for a nuclear power plant, development of additional guidance and training could improve the likelihood that the operating crew would implement effective sever-accident management measures. The US Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe-accident research results to enhance the capability of the plant operating crew to effectively manage severe accidents. One element of this program includes identification of the information needed by the operating crew in severe-accident situations. This paper discusses a method developed for identifying these information needs and its application. The methodology has been applied to a generic reactor design representing a PWR with a large dry containment. The information needs were identified by systematically determining what information is needed to assess the health of the critical functions, identify the presence of challenges, select strategies, and assess the effectiveness of these strategies. This method allows the systematic identification of information needs for a broad range of severe-accident scenarios and can be validated by exercising the functional models for any specific event sequence.

Environmental consequences of releases from nuclear accidents

International Nuclear Information System (INIS)

Tveten, U.

1990-01-01

The primary purpose of this report is to present the results of a four-year Nordic cooperation program in the area of consequence assessment of nuclear accidents with large releases to the environment. This program was completed in 1989. Related information from other research programs has also been described, so that many chapters of the report reflect the current status in the respective areas, in addition to containing the results of the Nordic program. (author) 179 refs

[Research on accidents in a tire-producing plant].

Science.gov (United States)

Mete, R; Sabatucci, A

1989-09-30

In the autumn of 1987 the U.S.L. health service (prevention, hygiene and occupational safety section) began a study about the accidents in a firm manufacturing tyres, placed in its own area. The retrospective enquiry starts from the analysis of typology, diffusion and seriousness of occupational accidents. The firm's accident register has been analyzed and integrated with other necessary information provided by the firm, by I.N.A.I.L. and by the air force metereological service. The study has been carried out on data concerning the following years: 1984-1985-1986. The accidents considered, implied absence from work and were divided as follows: for absence up till 3 days (in franchise), and more than 3 days (indemnified), applying the average value calculated on one year of the three analyzed. Every accident has been analyzed per year, month, day, hour of event. According to the classes: circumstances, kind of lesion, site of lesion, period of absence from work. The indices of: frequency, seriousness, incidence, mean duration have been calculated. The average monthly values of temperature: max and min. of the area and to the average monthly amount of processed elastomer (rate of production). The statistics we obtained, justified the study and showed the operative solution. The aspect of sanitary education and the general psychological aspect regarding the accident have been considered. Moreover the general operative solutions for the firm and specific ones for every department and for every position have been shown and faced up to. In this way, according to the risks that have emerged from the enquiries on previous accidents and thanks to direct inspection. it was possible to prevent accidents.

Analysis of avalanche risk factors in backcountry terrain based on usage frequency and accident data in Switzerland

Science.gov (United States)

Techel, F.; Zweifel, B.; Winkler, K.

2015-09-01

Recreational activities in snow-covered mountainous terrain in the backcountry account for the vast majority of avalanche accidents. Studies analyzing avalanche risk mostly rely on accident statistics without considering exposure (or the elements at risk), i.e., how many, when and where people are recreating, as data on recreational activity in the winter mountains are scarce. To fill this gap, we explored volunteered geographic information on two social media mountaineering websites - bergportal.ch and camptocamp.org. Based on these data, we present a spatiotemporal pattern of winter backcountry touring activity in the Swiss Alps and compare this with accident statistics. Geographically, activity was concentrated in Alpine regions relatively close to the main Swiss population centers in the west and north. In contrast, accidents occurred equally often in the less-frequented inner-alpine regions. Weekends, weather and avalanche conditions influenced the number of recreationists, while the odds to be involved in a severe avalanche accident did not depend on weekends or weather conditions. However, the likelihood of being involved in an accident increased with increasing avalanche danger level, but also with a more unfavorable snowpack containing persistent weak layers (also referred to as an old snow problem). In fact, the most critical situation for backcountry recreationists and professionals occurred on days and in regions when both the avalanche danger was critical and when the snowpack contained persistent weak layers. The frequently occurring geographical pattern of a more unfavorable snowpack structure also explains the relatively high proportion of accidents in the less-frequented inner-alpine regions. These results have practical implications: avalanche forecasters should clearly communicate the avalanche danger and the avalanche problem to the backcountry user, particularly if persistent weak layers are of concern. Professionals and recreationists, on the

Fuel solution criticality accident studies with the SILENE reactor: phenomenology, consequences and simulated intervention

International Nuclear Information System (INIS)

Barbry, F.

1984-01-01

After defining the content and the objectives of criticality accident studies, the SILENE reactor, a means of studying fuel solution criticality accidents, is presented. Information obtained from the CRAC and SILENE experimental programs are then presented; they concern power excursion phenomenology, radiological consequences, and finally guide-lines for current and future programs

Gas-cooled fast reactor safety - and overview and status of the U.S. program

International Nuclear Information System (INIS)

Torri, A.; Buttemer, D.R.

1981-01-01

In the revised GCFR Safety Program Plan a quantitative risk limit line has been adopted to establish requirements for the safety related functions and systems. The risk limit line is derived from an interpretation of NRC established licensing requirements, including those for LMFBR's. Multiple barriers to the progression of accident sequences are defined in the form of six Lines of Protection (LOPs). LOPs-1 to 3 are dedicated to accident prevention and represent the normal operating systems, the dedicated safety systems and the inherent design features, respectively. LOPs-4 to 6 are dedicated to the mitigation of core melt accident consequences and include in-vessel accident containment, secondary containment integrity and radiological attenuation, respectively. Cumulative frequency limits and consequence limits are established for each LOP. Design features associated with each LOP are described and the results of supporting safety analyses are summarized. (author)

Dose estimates in a loss of lead shielding truck accident.

Energy Technology Data Exchange (ETDEWEB)

Dennis, Matthew L.; Osborn, Douglas M.; Weiner, Ruth F.; Heames, Terence John (Alion Science & Technology Albuquerque, NM)

2009-08-01

The radiological transportation risk & consequence program, RADTRAN, has recently added an updated loss of lead shielding (LOS) model to it most recent version, RADTRAN 6.0. The LOS model was used to determine dose estimates to first-responders during a spent nuclear fuel transportation accident. Results varied according to the following: type of accident scenario, percent of lead slump, distance to shipment, and time spent in the area. This document presents a method of creating dose estimates for first-responders using RADTRAN with potential accident scenarios. This may be of particular interest in the event of high speed accidents or fires involving cask punctures.

Updating of adventitious fuel pin failure frequency in sodium-cooled fast reactors and probabilistic risk assessment on consequent severe accident in Monju

International Nuclear Information System (INIS)

Fukano, Yoshitaka; Kurisaka, Kenichi; Nishimura, Masahiro; Naruto, Kenichi

2015-01-01

Experimental studies, deterministic approaches and probabilistic risk assessments (PRAs) on local fault (LF) propagation in sodium-cooled fast reactors (SFRs) have been performed in many countries because LFs have been historically considered as one of the possible causes of severe accidents. Adventitious-fuel-pin-failures (AFPFs) have been considered to be the most dominant initiators of LFs in these PRAs because of their high frequency of occurrence during reactor operation and possibility of fuel-element-failure-propagation (FEFP). A PRA on FEFP from AFPF (FEFPA) in the Japanese prototype SFR (Monju) was performed in this study based on the state-of-the-art knowledge, reflecting the most recent operation procedures under off-normal conditions. Frequency of occurrence of AFPF in SFRs which was the initiating event of the event tree in this PRA was updated using a variety of methods based on the above-mentioned latest review on experiences of this phenomenon. As a result, the frequency of occurrence of, and the core damage frequency (CDF) from, AFPF in Monju was significantly reduced to a negligible magnitude compared with those in the existing PRAs. It was, therefore concluded that the CDF of FEFPA in Monju could be comprised in that of anticipated transient without scram or protected loss of heat sink events from both the viewpoint of occurrence probability and consequences. (author)

Accident management insights after the Fukushima Daiichi NPP accident

International Nuclear Information System (INIS)

Degueldre, Didier; Viktorov, Alexandre; Tuomainen, Minna; Ducamp, Francois; Chevalier, Sophie; Guigueno, Yves; Tasset, Daniel; Heinrich, Marcus; Schneider, Matthias; Funahashi, Toshihiro; Hotta, Akitoshi; Kajimoto, Mitsuhiro; Chung, Dae-Wook; Kuriene, Laima; Kozlova, Nadezhda; Zivko, Tomi; Aleza, Santiago; Jones, John; McHale, Jack; Nieh, Ho; Pascal, Ghislain; ); Nakoski, John; Neretin, Victor; Nezuka, Takayoshi; )

2014-01-01

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

A PC based multi-CPU severe accident simulation trainer

International Nuclear Information System (INIS)

Jankowski, M.W.; Bienarz, P.P.; Sartmadjiev, A.D.

2004-01-01

MELSIM Severe Accident Simulation Trainer is a personal computer based system being developed by the International Atomic Energy Agency and Risk Management Associates, Inc. for the purpose of training the operators of nuclear power stations. It also serves for evaluating accident management strategies as well as assessing complex interfaces between emergency operating procedures and accident management guidelines. The system is being developed for the Soviet designed WWER-440/Model 213 reactor and it is plant specific. The Bohunice V2 power station in the Slovak Republic has been selected for trial operation of the system. The trainer utilizes several CPUs working simultaneously on different areas of simulation. Detailed plant operation displays are provided on colour monitor mimic screens which show changing plant conditions in approximate real-time. Up to 28 000 curves can be plotted on a separate monitor as the MELSIM program proceeds. These plots proceed concurrently with the program, and time specific segments can be recalled for review. A benchmarking (limited in scope) against well validated thermal-hydraulic codes and selected plant accident data (WWER-440/213 Rovno NPP, Ukraine) has been initiated. Preliminary results are presented and discussed. (author)

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

International Nuclear Information System (INIS)

Takahashi, Fumiaki; Shigemori, Yuji; Seki, Akiyuki

2009-07-01

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

Emergency response plan for accidents in Saudi Arabia

International Nuclear Information System (INIS)

Al-Solaiman, K.M.; Al-Arfaj, A.M.; Farouk, M.A.

2000-01-01

This paper presents a brief description of the general emergency plan for accidents involving radioactive materials in the Kingdom of Saudi Arabia. Uses of radioactive materials and radiation sources and their associated potential accident are specified. Most general accident scenarios of various levels have been determined. Protective measures have been specified to reduce individual and collective doses arising during accident situations. Intervention levels for temporary exposure situations, as established in the IAEA's basic safety standards for protection against ionising radiation and for the safety of radiation sources, are adopted as national intervention levels. General procedures for implementation of the response plan, including notification and radiological monitoring instrumentation and equipment, are described and radiation monitoring teams are nominated. Training programs for the different parties which may be called upon to respond are studied and will be started. (author)

Recommendations for prevention of radiation accident in industrial gammagraphy

International Nuclear Information System (INIS)

Souza, L.S.; Silva, F.C.A. da

2017-01-01

Industrial Gammagraphy plays an important role in the quality control of various materials and components. It is classified by the International Atomic Energy Agency - IAEA as Category 2, due to its radiation risk caused by the use of high activity radioactive sources. This risk is based on the harmful consequences of human health, described in some accidents in the world, due to failures. In 2012, the 'Brazilian National Workshop on Accident Prevention in Industrial Gammagraphy' was carried out by DIAPI/CNEN, with the objective of disseminating knowledge about radiation accidents. At the time, the IRD/CNEN-RJ carried out a survey with the 75 participants using a form with 22 recommendations to prevent radiological accidents, in order to select the 10 most voted. A statistical study, using the 'Frequency Distribution' method, was performed to define 10 recommendations. The percentage and vote results were obtained by category of the participants and the 10 most important recommendations were defined to prevent radiation accidents. The recommendation that came in first place was 'Always use an individual monitor with alarm during all work'

Comparative aspects of frequency of illness in agricultural workers from the zoned of intensive radiation contamination (after Chernobyl accident)

International Nuclear Information System (INIS)

Faktorov, Yi.Je.; Vyitte, P.M.; Jershova, M.O.; Belemets', M.P.; Chebanova, O.V.

1997-01-01

To study the frequency of illness with temporary loss of working ability (FITLWA) in the persons of main agricultural professions (drives, field workers, cattle breeders) working in the zone of strict radiation control (Polissia District, Kyiv Region) when compared with the 'clean' Lokhvytsia District, Poltava Region (the controls). FITLWA values were studied in 1496 workers. Within four year period after the accident at CAPP the residents of Polissia District received the dose of 2 rem due to Cesium-137, 134. Besides they were irradiated by incorporated Strontium-90, 89 and Plutonium-238,239,240. FITLWA values (duration of the illness, the number of patients) in all professional groups from this district have been established to be higher than in the controls. First of all it concerned the diseases of peripheral nervous system, gastrointestinal tract, respiratory organs, mental disorders, cardiovascular pathology

Source term and radiological consequences of the Chernobyl accident

International Nuclear Information System (INIS)

Mourad, R.

1987-09-01

This report presents the results of a study of the source term and radiological consequences of the Chernobyl accident. The results two parts. The first part was performed during the first 2 months following the accident and dealt with the evaluation of the source term and an estimate of individual doses in the European countries outside the Soviet Union. The second part was performed after August 25-29, 1986 when the Soviets presented in a IAEA Conference in Vienna detailed information about the accident, including source term and radiological consequences in the Soviet Union. The second part of the study reconfirms the source term evaluated in the first part and in addition deals with the radiological consequences in the Soviet Union. Source term and individual doses are calculated from measured post-accident data, reported by the Soviet Union and European countries, microcomputer program PEAR (Public Exposure from Accident Releases). 22 refs

Criticality accident studies and methodology implemented at the CEA

International Nuclear Information System (INIS)

Barbry, Francis; Fouillaud, Patrick; Reverdy, Ludovic; Mijuin, Dominique

2003-01-01

Based on the studies and results of experimental programs performed since 1967 in the CRAC, then SILENE facilities, the CEA has devised a methodology for criticality accident studies. This methodology integrates all the main focuses of its approach, from criticality accident phenomenology to emergency planning and response, and thus includes aspects such as criticality alarm detector triggering, airborne releases, and irradiation risk assessment. (author)

The need to study of bounding accident in reprocessing plant

International Nuclear Information System (INIS)

Segawa, Satoshi; Fujita, Kunio

2013-01-01

There is a clear consensus that the severe accident corresponds to the core damage accident for power reactors. On the other hand, for FCFs, there is no clear consensus on what is the accident to assess the safety in the region of beyond design basis, or what is the accident which has very low probability but large consequence. The need to examine a bounding consequence of each type of accident is explained to advance the rationality of safety management and regulation and, as a result, to reinforce the safety of a reprocessing plant. The likelihood of occurrence of an accident causing a bounding consequence should correspond to that of a severe accident at a nuclear power plant. The bounding consequence will be derived using the deterministic method and sound engineering judgment supplemented by the probabilistic method. Once an agreement on such a concept is reached among regulators, operators and related experts it will help to provide a solid basis to ensure the safety of a reprocessing plant independent of that of a nuclear power plant. In this paper, we show a preliminary risk profile of RRP calculated by QSA (Quantitative Safety Assessment) which JNFL developed. The profile shows that bounding consequences of various accidents in a range of occurrence frequency corresponding to a severe accident at a nuclear power plant. And we find that the bounding consequence of high-level liquid waste boiling is the largest among all in this range. Therefore, the risk of this event is shown in this paper as an example. To build a common consensus about bounding accidents among concerned parties will encourage regulatory body to introduce such an idea for more effective regulation with scientific rationality. Additionally the study of bounding accidents can contribute to substantial development for accident management strategy as reprocessing operators. (authors)

Unavoidable Accident

OpenAIRE

Grady, Mark F.

2009-01-01

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

Examining accident reports involving autonomous vehicles in California

Science.gov (United States)

Nader, Nazanin; Eurich, Sky O.; Tripp, Michelle; Varadaraju, Naresh

2017-01-01

Autonomous Vehicle technology is quickly expanding its market and has found in Silicon Valley, California, a strong foothold for preliminary testing on public roads. In an effort to promote safety and transparency to consumers, the California Department of Motor Vehicles has mandated that reports of accidents involving autonomous vehicles be drafted and made available to the public. The present work shows an in-depth analysis of the accident reports filed by different manufacturers that are testing autonomous vehicles in California (testing data from September 2014 to March 2017). The data provides important information on autonomous vehicles accidents’ dynamics, related to the most frequent types of collisions and impacts, accident frequencies, and other contributing factors. The study also explores important implications related to future testing and validation of semi-autonomous vehicles, tracing the investigation back to current literature as well as to the current regulatory panorama. PMID:28931022

A statistical description of the types and severities of accidents involving tractor semi-trailers

International Nuclear Information System (INIS)

Clauss, D.B.; Wilson, R.K.; Blower, D.F.; Campbell, K.L.

1994-06-01

This report provides a statistical description of the types and severities of tractor semi-trailer accidents involving at least one fatality. The data were developed for use in risk assessments of hazardous materials transportation. Several accident databases were reviewed to determine their suitability to the task. The TIFA (Trucks Involved in Fatal Accidents) database created at the University of Michigan Transportation Research Institute was extensively utilized. Supplementary data on collision and fire severity, which was not available in the TIFA database, were obtained by reviewing police reports for selected TIFA accidents. The results are described in terms of frequencies of different accident types and cumulative distribution functions for the peak contact velocity, rollover skid distance, fire temperature, fire size, fire separation, and fire duration

Characteristics of domestic accidents and related factors in Karatas district of Adana province, Turkey

Directory of Open Access Journals (Sweden)

Muhsin Akbaba

2015-06-01

Full Text Available Background: Domestic accidents (DA are an important, but still neglected public health problem in Turkey. We aimed to determine the prevalence of DA, their physical and residential characteristics, including affecting factors, in the centre of semi-rural Mediterranean District Karatas of Turkey. Methods: This cross-sectional descriptive study, performed between October 2011 and April 2012 with face-to-face interviews by the researcher physician, collected DA data about 569 people living in 141 houses, selected by simple random sampling method, among 2726 houses. Results: DA frequency was 45.4% and 30.2% in terms of houses included in the study and the whole population resident in those houses, respectively. DA most frequently occurred in kitchen (38.9%, affecting mostly upper extremities (45.9%. Most of the injured, i.e.124(72.1% people didn't apply to any health institution; 43(25.0% people were treated at outpatient departments. Only 5(2.9% required hospitalization. None of the accidents resulted in death. The season with the highest frequency of accidents was found to be the summer. DA frequency was positively proportional to the house-hold size (p0.05. Conclusion: Primary health service providing institutions can be suggested for the determination of risk factors for DA in addition to the establishment of a national accidents surveillance system. [TAF Prev Med Bull 2015; 14(3.000: 229-234

Real and mythical consequences of Chernobyl accident

International Nuclear Information System (INIS)

Osmachkin, V.S.

1999-01-01

This presentation describes the public Unacceptance of Nuclear Power as a consequence of Chernobyl Accident, an accident which was a severest event in the history of the nuclear industry. It was a shock for everybody, who has been involved in nuclear power programs. But nobody could expect that it was also the end romantic page in the nuclear story. The scale of the detriment was a great, and it could be compared with other big technological man-made catastrophes. But immediately after an accident mass media and news agencies started to transmit an information with a great exaggerations of the consequences of the event. In a report on the Seminar T he lessons of the Chernobyl - 1' in 1996 examples of such incorrect information, were cited. Particularly, in the mass media it was declared that consequences of the accident could be compared with a results of the second world war, the number of victims were more than hundred thousand people, more than million of children have the serious health detriments. Such and other cases of the misconstruction have been called as myths. The real consequences of Chernobyl disaster have been summed on the International Conference 'One decade after Chernobyl' - 2, in April 1996. A very important result of the Chernobyl accident was a dissemination of stable unacceptance of the everything connected with 'the atom'. A mystic horror from invisible mortal radiation has been inspired in the masses. And from such public attitude the Nuclear Power Programs in many countries have changed dramatically. A new more pragmatic and more careful atomic era started with a slogan: 'Kernkraftwerk ? Nein, danke'. No doubt, a Chernobyl accident was a serious technical catastrophe in atomic industry. The scale of detriment is connected with a number of involved peoples, not with a number of real victims. In comparison with Bhopal case, earthquakes, crashes of the airplanes, floods, traffic accidents and other risky events of our life - the Chernobyl is

Using MARS to assist in managing a severe accident

International Nuclear Information System (INIS)

Raines, J.C.; Hammersley, R.J.; Henry, R.E.

2004-01-01

During an accident, information about the current and possible future states of the plant provides guidance for accident managers in evaluating which actions should be taken. However, depending upon the nature of the accident and the stress levels imposed on the plant staff responding to the accident the current and future plant assessments may be very difficult or nearly impossible to perform without supplemental training and/or appropriate tools. The MAAP Accident Response System (MARS) has been developed as a calculational aid to assist the responsible accident management individuals. Specifically MARS provides additional insights on the current and possible future states of the plant during an accident including the influence of operator actions. In addition to serving as a calculational aid, the MARS software can be an effective means for providing supplemental training. The MARS software uses engineering calculations to perform an integral assessment of the plant status including a consistency assessment of the available instrumentation. In addition, it uses the Modular Accident Analysis Program (MAAP) to provide near term predictions of the plant response if corrective actions are taken. This paper will discuss the types of information that are beneficial to the accident manager and how MARS addresses each. The MARS calculational functions include: instrumentation, validation and simulation, projected operator response based on the EOPs, as well as estimated timing and magnitude of in-plant and off-site radiation dose releases. Each of these items is influential in the management of a severe accident. (author)

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...

Accidents, 'black swans' and risks

International Nuclear Information System (INIS)

Luxat, J.C.

2013-01-01

Major accidents and natural disasters with severe consequences have occurred in all sectors of industrial activity with relatively high frequency. The severe consequences of concern involve either significant loss of life or major economic loss, or both loss of life and economic loss. Such events have the last two years been referred to as 'black swan' events following publication of Taleb's bestselling book. These events demonstrate limits to PRA application that arise from the underlying high uncertainty associated with the estimation of frequency of occurrence of such events. An approach is proposed in this paper that, consistent with the concept of defense in depth employed by the nuclear industry, augments probabilistic risk assessment with a methodology based upon 'threat - risk assessment'. This approach shifts these very low frequency high consequence 'black swan' events out of the probabilistic risk assessment domain into a deterministic emergency response assessment domain. (author)

Preliminary neutronic assessment for ATF (Accident Tolerant Fuel) based on iron alloy

International Nuclear Information System (INIS)

Abe, Alfredo; Carluccio, Thiago; Piovezan, Pamela; Giovedi, Claudia; Martins, Marcelo R.

2015-01-01

After Fukushima Daiichi nuclear accident in 2011, the nuclear fuel performance under accident condition became a very important issue and currently different research and development program are in progress toward to reliability and withstand under accident condition. These initiatives are known as ATF (Accident Tolerant Fuel) R and D program, which many countries with different research institutes, fuel vendors and others are nowadays involved. Accident Tolerant Fuel (ATF) can be defined as enhanced fuel which can tolerate loss of active cooling system capability for a considerably longer time period and the fuel/cladding system can be maintained without significant degradation and can also improve the fuel performance during normal operations and transients, as well as design-basis accident (DBA) and beyond design-basis (BDBA) accident. Different materials have being proposed as fuel cladding candidates considering thermo-mechanical properties and lower reaction kinetic with steam and slower hydrogen production. The aim of this work is to perform a neutronic assessment for several cladding candidates based on iron alloy considering a standard PWR fuel rod (fuel pellet and dimension). The purpose of the assessment is to address different parameters that might contribute for possible neutronic reactivity gain in order to overcome the penalty due to increase of neutron absorption in the cladding materials. All the neutronic assessment is performed using MCNP, Monte Carlo code. (author)

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

International Nuclear Information System (INIS)

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

2004-01-01

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

An Epidemiological Study of Road Traffic Accidents in Guilan Province, Northern Iran in 2012

Science.gov (United States)

Mohtasham-Amiri, Zahra; Dastgiri, Saeed; Davoudi-kiakalyeh, Ali; Imani, Ali; Mollarahimi, Keyvan

2016-01-01

Objective: To determine the epidemiological characteristics of the road traffic injuries (RTIs) in Guilan province, northern Iran. Methods: This study was a cross-sectional study which included all of RTIs admitted to medical centers of Guilan province (northern Iran) during 2012. ICD-10 was used as diagnostic criteria. Demographic variables also injury circumstance and in hospital variables such as length of stay, time of admission, type of surgery, ICU admission, final outcome and mechanism of injury, anatomical part of injury according to Abbreviated Injury Scale (AIS) classification were derives from records by trained research team. Descriptive data is reported. The predictors of mortality were also determined. Results: The prevalence of road traffic injuries in Guilan province was 31 in 10,000 populations. Of total 7671 accidents, 5976 (77.9%) were men and 1695 (22.1%) were women. Mean age of these victims was 33.3 ± 17.289 years (32.64±16.939 for men, 35.62±18.312 for women). Most of them (32.5%) were 20-29 years old. Motorcycle-car accidents had the highest frequency followed by car-car crashes and car accidents involving pedestrians. Most of the patients (85.9%) were hospitalized and 280 injured died (3.7%). Upper extremities were the most sites of injuries. Male sex, length of hospital stay, multiple injuries and increased age were associated with road traffic accident associated mortality. Conclusion: RTIs cause enormous death and disability in this area and more road traffic preventive programs should be enforcement in these areas to reduce incidences RTI. PMID:27878129

Main lessons based on the Chernobyl nuclear power plant accident liquidation experience

International Nuclear Information System (INIS)

Vasil'chenko, V.N.; Nosovskij, A.V.

2006-01-01

The authors review the main lessons of the Chernobyl nuclear power plant accident and the liquidation of its consequences in the area of the nuclear reactors safety operation, any major accident management, liquidation accident consequences criteria, emergency procedures, preventative measures and treatment irradiated victims, the monitoring methods etc. The special emphasis is put on the questions of the emergency response and the antiaccidental measures planning in frame of international cooperation program

Stress in accident and post-accident management at Chernobyl

International Nuclear Information System (INIS)

Girard, P.; Dubreuil, G.H.

1996-01-01

The effects of the Chernobyl nuclear accident on the psychology of the affected population have been much discussed. The psychological dimension has been advanced as a factor explaining the emergence, from 1990 onwards, of a post-accident crisis in the main CIS countries affected. This article presents the conclusions of a series of European studies, which focused on the consequences of the Chernobyl accident. These studies show that the psychological and social effects associated with the post-accident situation arise from the interdependency of a number of complex factors exerting a deleterious effect on the population. We shall first attempt to characterise the stress phenomena observed among the population affected by the accident. Secondly, we will be presenting an anlysis of the various factors that have contributed to the emerging psychological and social features of population reaction to the accident and in post-accident phases, while not neglecting the effects of the pre-accident situation on the target population. Thirdly, we shall devote some initial consideration to the conditions that might be conducive to better management of post-accident stress. In conclusion, we shall emphasise the need to restore confidence among the population generally. (Author)

The survey of occupational accidents in Yazd gas agency (2013

Directory of Open Access Journals (Sweden)

Amir Hossein Khoshakhlagh

2016-10-01

Full Text Available Background: Existence of coordinated and professional safety system to prevent occurrence of accidents and potential hazards seem to be essential in installing networks of gas distribution projects. Objective: To survey work-related accidents and safety performance indices in project implementation unit of Yazd gas agency. Methods: This analytical study was conducted on 197 of workforce in Yazd gas agency in 2013 that were selected by census and they were male. Demographic and accident information were gathered using a self-made questionnaire and face- to- face interview, and required information obtained from dossier to determine the safety performance indicators. Safety performance indicators were calculated in separately of 13 types occupations in project implementation unit of gas agency and data were analyzed using T-test. Findings: The highest accident frequency and severity rate were related to digging occupation and then metal line welding. Consequences of accidents were cuts (%56.7 and soreness (%14.9. The causes of accidents were related to uselessness of personal protective equipment (%25.2 and lack of precision in the task (%19.3. The highest rate of accident was observed among the age group 20-29 years with work experience of 4-6 years. Conclusion: According to the findings of this study and the risk of gas processes, it seems to be necessary the implementation of integrated management systems and training of workers about safety rules to improve the safety culture and prevent accidents.

Benchmarking MARS (accident management software) with the Browns Ferry fire

International Nuclear Information System (INIS)

Dawson, S.M.; Liu, L.Y.; Raines, J.C.

1992-01-01

The MAAP Accident Response System (MARS) is a userfriendly computer software developed to provide management and engineering staff with the most needed insights, during actual or simulated accidents, of the current and future conditions of the plant based on current plant data and its trends. To demonstrate the reliability of the MARS code in simulatng a plant transient, MARS is being benchmarked with the available reactor pressure vessel (RPV) pressure and level data from the Browns Ferry fire. The MRS software uses the Modular Accident Analysis Program (MAAP) code as its basis to calculate plant response under accident conditions. MARS uses a limited set of plant data to initialize and track the accidnt progression. To perform this benchmark, a simulated set of plant data was constructed based on actual report data containing the information necessary to initialize MARS and keep track of plant system status throughout the accident progression. The initial Browns Ferry fire data were produced by performing a MAAP run to simulate the accident. The remaining accident simulation used actual plant data

Vaporization of structural materials in severe accidents

International Nuclear Information System (INIS)

Lorenz, R.A.

1982-01-01

Vaporized structural materials form the bulk of aerosol particles that can transport fission products in severe LWR accidents. As part of the Severe Accident Sequence Analysis (SASA) program at Oak Ridge National Laboratory, a model has been developed based on a mass transport coefficient to describe the transport of materials from the surface of a molten pool. In many accident scenarios, the coefficient can be calculated from existing correlations for mass transfer by natural convection. Data from SASCHA fuel melting tests (Karlsruhe, Germany) show that the partial pressures of many of the melt components (Fe, Cr, Co, Mn, Sn) required for the model can be calculated from the vapor pressures of the pure species and Raoult's law. These calculations indicate much lower aerosol concentrations than reported in previous studies

Assessment of CRBR core disruptive accident energetics

International Nuclear Information System (INIS)

Theofanous, T.G.; Bell, C.R.

1984-03-01

The results of an independent assessment of core disruptive accident energetics for the Clinch River Breeder Reactor are presented in this document. This assessment was performed for the Nuclear Regulatory Commission under the direction of the CRBR Program Office within the Office of Nuclear Reactor Regulation. It considered in detail the accident behavior for three accident initiators that are representative of three different classes of events; unprotected loss of flow, unprotected reactivity insertion, and protected loss of heat sink. The primary system's energetics accommodation capability was realistically, yet conservatively, determined in terms of core events. This accommodation capability was found to be equivalent to an isentropic work potential for expansion to one atmosphere of 2550 MJ or a ramp rate of about 200 $/s applied to a classical two-phase disassembly

How to Reduce Those Costly School Bus Accidents.

Science.gov (United States)

Farmer, Ernest

1985-01-01

The deterrent to school bus accidents is preparedness. Training programs for drivers and mechanics, equipment specifications, and a community support base are some of the ways to prevent tragedy. (MLF)

Fatal accidents in nighttime vs. daytime highway construction work zones.

Science.gov (United States)

Arditi, David; Lee, Dong-Eun; Polat, Gul

2007-01-01

Awareness about worker safety in nighttime construction has been a major concern because it is believed that nighttime construction creates hazardous work conditions. However, only a few studies provide valuable comparative information about accident characteristics of nighttime and daytime highway construction activities. This study investigates fatal accidents that occurred in Illinois highway work zones in the period 1996-2001 in order to determine the safety differences between nighttime and daytime highway construction. The lighting and weather conditions were included into the study as control parameters to see their effects on the frequency of fatal accidents occurring in work zones. According to this study, there is evidence that nighttime construction is more hazardous than daytime construction. The inclusion of a weather parameter into the analysis has limited effect on this finding. The study justifies establishing an efficient work zone accident reporting system and taking all necessary measures to enhance safety in nighttime work zones.

A Content Analysis of News Media Coverage of the Accident at Three Mile Island.

Science.gov (United States)

Stephens, Mitchell; Edison, Nadyne G.

A study was conducted for the President's Commission on the Accident at Three Mile Island to analyze coverage of the accident by ten news organizations: two wire services, three commercial television networks, and five daily newspapers. Copies of all stories and transcripts of news programs during the first week of the accident were examined from…

Frequency Calculation For Loss Coolant Accident In The Nuclear Reactor

International Nuclear Information System (INIS)

Sony, DT

1996-01-01

LOCA as initiating event is engineering judgement, because it is rare condition. So, to determine LOCA frequency used be probability and statistic method. By probability and statistic method was estimated from size, weld, age, learning curve and quality, etc. it has been calculated for LOCA frequency in the simplified piping system model, especially estimates from size and weld factors. From calculation, LOCA frequency is 9,82.10 - 6/year

Hygienic training of population being victims of the Chernobyl accident

International Nuclear Information System (INIS)

Terman, A.V.; Mozgovaya, A.V.; Polesskij, V.A.

1995-01-01

Study results on the role of social factors in formation of attitude to own health and its self-evaluation by the population of the regions, subjected to impact of the Chernobyl NPP accident. An extremely important component block is determined in the programs on hygienic training of the population being victims of the accident, namely, adequate information of the public on dose-effect dependencies, on radionuclide behaviour in the environmental objects, on possible measures for reduction of undesirable effects. Necessity is noted of transfer from universal programs of hygienic training to differential ones up to individual training

Assessment of the accident response of a light-water-moderated breeder-reactor system: AWBA development program

International Nuclear Information System (INIS)

High, H.M.

1983-05-01

The predicted accident response for a light water moderated, thorium/U-233 fueled, seed-blanket reactor concept was assessed. The first part of the assessment compared breeder accident response with that of a current commercial pressurized water reactor design for several different types of transients. Based on these comparisons and a review of the various parameter differences between the breeder and a U-235 fueled plant, the second part of the assessment studied the breeder accident behavior in more detail, particularly in areas of potential concern. Based on the two parts of the assessment, it was concluded that the breeder accident response would be very similar to that of present commercial pressurized water reactor plants. The large Doppler and moderator reactivity coefficients of the breeder would significantly reduce the severity of many of the accidents that must be considered. It is expected that the accident response of the breeder can be shown to meet regulatory criteria

Accident Diagnosis and Prognosis Aide (ADPA)

International Nuclear Information System (INIS)

Gunter, A.D.; Touchton, R.A.

1987-01-01

This presentation provides a demonstration of a prototypical expert system developed by Technology Applications, Inc. (TAI) under a contract with the Department of Energy as a part of their Small Business Innovation Research Program. The Accident Diagnosis and Prognosis Aide (ADPA) Demonstration Prototype is a working scale model of a real-time expert system which: Diagnoses an accident situation (as well as a number of underlying failures, events, and conditions deduced along the way). Calculates the change in the likelihood of core damage as a function of the events and failures diagnosed. Dynamically generates a recovery procedure tailored to the specific plant state at hand

Investigation report on causes of radiation underexposure accident at Yamagata University Hospital and Prevention of Similar accident

International Nuclear Information System (INIS)

2005-01-01

The accident in the title was announced on February 18, 2004 by the hospital, which asked its investigation immediately. The group based on 4 academic societies concerned, thereby started investigations of the in-house reports on the accident and of subsequent hospital visit in March, which involved hearing from personnel concerned, physical/technological examinations and clinical evaluation, with respect to the hospital system for radiation treatment, flow of the treatment, accident details, estimation of the actual expose dose and classification of patients. The investigational group found for the actual number of patients underexposed to be 36 (63, in the in-house report) in 1,377. The cause of the accident was thought essentially the input error for the correct power coefficient 1.032 to be a wrong one 1.320 for 15 x 15 cm 4 MV X-ray. The error had been overlooked by the contract operator from the introduction of the treatment planning equipment in 1999. For prevention, setting up of quality assurance (QA) program by the hospital, the user itself, was pointed out necessary. Making the guideline for introducing the new equipment was conceivably an important work of the trader. (N.I.)

CiOpt: a program for optimization of the frequency response of linear circuits

OpenAIRE

Miró Sans, Joan Maria; Palà Schönwälder, Pere

1991-01-01

An interactive personal-computer program for optimizing the frequency response of linear lumped circuits (CiOpt) is presented. CiOpt has proved to be an efficient tool in improving designs where the inclusion of more accurate device models distorts the desired frequency response, as well as in device modeling. The outputs of CiOpt are the element values which best match the obtained and the desired frequency response. The optimization algorithms used (the Fletcher-Powell and Newton's methods,...

Probabilistic studies of accident sequences

International Nuclear Information System (INIS)

Villemeur, A.; Berger, J.P.

1986-01-01

For several years, Electricite de France has carried out probabilistic assessment of accident sequences for nuclear power plants. In the framework of this program many methods were developed. As the interest in these studies was increasing and as adapted methods were developed, Electricite de France has undertaken a probabilistic safety assessment of a nuclear power plant [fr

Simulation of severe accident using March-3 computer code

International Nuclear Information System (INIS)

Fernandes, A.; Nakata, H.

1991-01-01

The severe accident sensitivity analysis utilizing the March-3 approximate modelization options has been performed. The reference results against which the present results have been compared were obtained from the best published results for the most representative accident sequences: TMLU, S sub(2)DC sub(r) and S sub(2)DCF sub(r) for the Zion-1 reactor. The results of the present sensitivity analysis revealed the presence of very crude modelizations, in the March-3 program, to represent the critical phenomenologies involved in the severe accident sequences considered, even though large uncertainties must still be taken into account due primarily to the scarcity of the integral benchmark data. (author)

Introduction of the Space Shuttle Columbia Accident, Investigation Details, Findings and Crew Survival Investigation Report

Science.gov (United States)

Chandler, Michael

2010-01-01

As the Space Shuttle Program comes to an end, it is important that the lessons learned from the Columbia accident be captured and understood by those who will be developing future aerospace programs and supporting current programs. Aeromedical lessons learned from the Accident were presented at AsMA in 2005. This Panel will update that information, closeout the lessons learned, provide additional information on the accident and provide suggestions for the future. To set the stage, an overview of the accident is required. The Space Shuttle Columbia was returning to Earth with a crew of seven astronauts on 1Feb, 2003. It disintegrated along a track extending from California to Louisiana and observers along part of the track filmed the breakup of Columbia. Debris was recovered from Littlefield, Texas to Fort Polk, Louisiana, along a 567 statute mile track; the largest ever recorded debris field. The Columbia Accident Investigation Board (CAIB) concluded its investigation in August 2003, and released their findings in a report published in February 2004. NASA recognized the importance of capturing the lessons learned from the loss of Columbia and her crew and the Space Shuttle Program managers commissioned the Spacecraft Crew Survival Integrated Investigation Team (SCSIIT) to accomplish this. Their task was to perform a comprehensive analysis of the accident, focusing on factors and events affecting crew survival, and to develop recommendations for improving crew survival, including the design features, equipment, training and procedures intended to protect the crew. NASA released the Columbia Crew Survival Investigation Report in December 2008. Key personnel have been assembled to give you an overview of the Space Shuttle Columbia accident, the medical response, the medico-legal issues, the SCSIIT findings and recommendations and future NASA flight surgeon spacecraft accident response training. Educational Objectives: Set the stage for the Panel to address the

Accident management information needs for a BWR with a MARK I containment

Energy Technology Data Exchange (ETDEWEB)

Chien, D.N.; Hanson, D.J. (EG and G Idaho, Inc., Idaho Falls, ID (USA))

1991-05-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs.

Accident management information needs for a BWR with a MARK I containment

International Nuclear Information System (INIS)

Chien, D.N.; Hanson, D.J.

1991-05-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, information needs during severe accidents have been evaluated for Boiling Water Reactors (BWRs) with MARK 1 containments. This evaluation was performed using a methodology that identifies plant information needs necessary for personnel to: (a) diagnose that an accident is in progress, (b) select and implement strategies to prevent or mitigate the accident, and (c) monitor the effectiveness of these strategies. The information needs and capabilities identified are intended to form a basis for more comprehensive information needs assessments. The assessments will be performed during the analysis and development of specific strategies, which will be used in accident management prevention and mitigation. 3 refs., 4 figs., 2 tabs

Self-reported accidents

DEFF Research Database (Denmark)

Møller, Katrine Meltofte; Andersen, Camilla Sloth

2016-01-01

The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals.......The main idea behind the self-reporting of accidents is to ask people about their traffic accidents and gain knowledge on these accidents without relying on the official records kept by police and/or hospitals....

Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

International Nuclear Information System (INIS)

1992-04-01

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

Proceedings of the Specialist Meeting on Severe Accident Management Programme Development

Energy Technology Data Exchange (ETDEWEB)

NONE

1992-04-15

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

Study on LOOP and SBO Frequency for Multi-Unit PSA

Energy Technology Data Exchange (ETDEWEB)

Jin, Kyung Ho; Heo, Gyun Young [Kyunghee University, Yongin (Korea, Republic of)

2016-05-15

In conventional single unit PSA, it was assumed that all accidents or events are independent and the risk of only one unit has been evaluated. In other words, the possibility that simultaneous events occur on multiple units was excluded because it was assumed that the probability of concurrent events were extremely low. After Fukushima accidents, however, it was found that external hazards such as tsunami may impact on multiple units; that means, for the sake of proper mitigation the risk of accidents in shared SSCs should be reevaluated. New risk metrics to improve conventional CDF (Core Damage frequency) based on reactor-year is needed to perform MUPSA (Multi-Unit Probabilistic Safety Assessment). IAEA suggested SCDF (Site CDF) as a risk metrics for MUPSA. The frequency based on reactor-year was converted to the frequency based on site-year. In addition, shared SSCs were modeled in single unit PSA as if those units have independent shared SSCs. Therefore, the risk of shared SSCs should be reevaluated. In this paper, the frequency of LOOP (Loss of Offsite Power), which is typically a multi-unit event, was evaluated and the frequency of SBO (Station Blackout) depending on LOOP frequency and emergency power systems such as EDG (Emergency Diesel Generator) and AAC (Alternate AC), that can mitigate SBO events, was modeled. This paper describes how to calculate LOOP and SBO frequency from the simple example for two-unit site with shared AAC. The events impacting on multiple units, which were excluded in conventional PSA, should be considered for MUPSA.

Accident Risks In The Energy Sector: Comparative Evaluations

International Nuclear Information System (INIS)

Hirschberg, S.; Burgherr, P.

2005-01-01

Severe accidents are considered one of the most controversial issues in current comparative studies of the environmental and health impact of energy systems. The present work focuses on severe accident scenarios relating to fossil energy chains (coal, oil and gas), nuclear power and hydro-power. The scope of the study is not limited to the power production (conversion) step of these energy chains, but, wherever applicable, also includes full energy chains. With the exception of the nuclear chain, the focus of the present work is on the evaluation of the historical experience of accidents. The basis for this evaluation is the comprehensive database ENSAD (Energy-Related Severe Accident Database), which has been established at PSI. For hypothetical nuclear accidents, a probabilistic technique has also been employed. The broader picture, derived from examination of full energy chains, leads, on a world-wide basis, to the conclusion that immediate fatality rates are much higher for the fossil chains than expected if only power plants are considered. Generally, immediate fatality rates are significantly higher for non-OECD countries than for OECD countries, and, in the case of hydro and nuclear, the difference is rather dramatic. In addition to aggregated values, frequency-consequence curves are also provided, since they not only reflect implicitly a ranking based on aggregated values, but also include such information as the observed, or predicted, chain-specific maximum extent of damages. Finally, damage and external costs of severe accidents for the different energy chains have been estimated, based on the unit cost values for the various consequence types. (author)

USNRC severe core damage assessment program

Energy Technology Data Exchange (ETDEWEB)

Hanson, J E [EG and G Idaho, Inc., Idaho Falls (USA); Johnston, W V; Kelber, C N [Nuclear Regulatory Commission, Washington, DC (USA)

1981-01-01

The accident at the Three Mile Island nuclear power station has significantly altered the perception of the importance of beyond-design-basis accidents in licensing and safety reviews of light-water reactors in the USA. Increased consideration will be given by the United States Nuclear Regulatory Commission to low-probability, high-risk core melt accidents in future licensing proceedings. To this end, the USNRC is mounting experimental and analytic methods development programs to provide the technical basis for future LWR design and licensing criteria related to class-9 accidents. The scope, objectives, and content of five major new programs addressing safety and licensing issues for beyond-design-basis accidents are reviewed and the rationale and logic for formulation of the programs is discussed.

The influence of the infrastructure characteristics in urban road accidents occurrence.

Science.gov (United States)

Vieira Gomes, Sandra

2013-11-01

This paper summarizes the result of a study regarding the creation of tools that can be used in intervention methods in the planning and management of urban road networks in Portugal. The first tool relates the creation of a geocoded database of road accidents occurred in Lisbon between 2004 and 2007, which allowed the definition of digital maps, with the possibility of a wide range of consultations and crossing of information. The second tool concerns the development of models to estimate the frequency of accidents on urban networks, according to different desegregations: road element (intersections and segments); type of accident (accidents with and without pedestrians); and inclusion of explanatory variables related to the road environment. Several methods were used to assess the goodness of fit of the developed models, allowing more robust conclusions. This work aims to contribute to the scientific knowledge of accidents phenomenon in Portugal, with detailed and accurate information on the factors affecting its occurrence. This allows to explicitly include safety aspects in planning and road management tasks. Copyright © 2013 Elsevier Ltd. All rights reserved.

Conclusions on severe accident research priorities

International Nuclear Information System (INIS)

Klein-Heßling, W.; Sonnenkalb, M.; Jacquemain, D.; Clément, B.; Raimond, E.; Dimmelmeier, H.; Azarian, G.; Ducros, G.; Journeau, C.; Herranz Puebla, L.E.; Schumm, A.; Miassoedov, A.; Kljenak, I.; Pascal, G.; Bechta, S.; Güntay, S.; Koch, M.K.; Ivanov, I.; Auvinen, A.; Lindholm, I.

2014-01-01

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

Summary of treat experiments on oxide core-disruptive accidents

International Nuclear Information System (INIS)

Dickerman, C.E.; Rothman, A.B.; Klickman, A.E.; Spencer, B.W.; DeVolpi, A.

1979-02-01

A program of transient in-reactor experiments is being conducted by Argonne National Laboratory in the Transient Reactor Test (TREAT) facility to guide and support analyses of hypothetical core-disruptive accidents (HCDA) in liquid-metal fast breeder reactors (LMFBR). Test results provide data needed to establish the response of LMFBR cores to hypothetical accidents producing fuel failure, coolant boiling, and the movement of coolant, molten fuel, and molten cladding. These data include margins to fuel failure, the modes of failure and movements, and evidence for identification of the mechanisms which determine the failure and movements. A key element in the program is the fast-neutron hodoscope, which detects fuel movement as a function of time during experiments

Strategies for the prevention and mitigation of severe accidents

International Nuclear Information System (INIS)

Ader, C.; Heusener, G.; Snell, V.G.

1999-01-01

The currently operating nuclear power plants have, in general, achieved a high level of safety, as a result of design philosophies that have emphasized concepts such as defense-in-depth. This type of an approach has resulted in plants that have robust designs and strong containments. These designs were later found to have capabilities to protect the public from severe accidents (accidents more severe than traditional design basis in which substantial damage is done to the reactor core). In spite of this high level of safety, it has also been recognized that future plants need to be designed to achieve an enhanced level of safety, in particular with respect to severe accidents. This has led both regulatory authorities and utilities to develop guidance and/or requirements to guide plant designers in achieving improved severe accident performance through prevention and mitigation. The considerable research programs initiated after the TMI-2 accident have provided a large body of technical data, analytical methods, and the expertise necessary to provide for an understanding of a range of severe accident phenomena. This understanding of the ways severe accidents can progress and challenge containments, combined with the wide use of probabilistic safety assessments, have provided designers of evolutionary water cooled reactors opportunities to develop designs that minimize the challenges to the plant and to the public from severe accidents, including the development of accident management strategies intended to further reduce the risk of severe accidents. This paper describes some of the recent progress made in the understanding of severe accidents and related safety assessment methodology and how this knowledge has supported the incorporation of features into representative evolutionary designs that will prevent or mitigate many of the severe accident challenges present in current plants. (author)

Safety culture and the accident at Three Mile Island

International Nuclear Information System (INIS)

Erp, Jan B. van

2002-01-01

Prior to the accident at Three Mile Island, little attention was being paid to the human role in the safe operation of civilian nuclear power plants. The investigation of the TMI accident showed that its root causes were primarily human-related. The Kemeny Report on the TMI accident does not use the term 'safety culture'; however, it fully identifies all relevant aspects of safety culture. It was only after the accident at Chernobyl that the term 'safety culture' came into widespread use. However, it should be noted that, during the years after TMI and before Chernobyl, already major changes had been instituted concerning human factors and human reliability in the civilian nuclear energy programs of many countries. Greater credit should be given to the remarkable insights developed by the Kemeny Commission as contained in the Kemeny Report. (author)

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

Directory of Open Access Journals (Sweden)

Robert P. Martin

2012-01-01

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

Analytical support for SAMG development as a part of accident management

International Nuclear Information System (INIS)

Honcarenko, R.

1999-01-01

The decision to built up and implement a comprehensive Accident Management Program applying best world-wide knowledge made during last year at Temelin. A small group of engineers dedicated to Accident Management was formed at Temelin NPP as a part of the plant organisation scheme. A short summary of these activities performed by this group is presented. (author)

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.

Nuclear power plant Severe Accident Research Plan

International Nuclear Information System (INIS)

Larkins, J.T.; Cunningham, M.A.

1983-01-01

The Severe Accident Research Plan (SARP) will provide technical information necessary to support regulatory decisions in the severe accident area for existing or planned nuclear power plants, and covers research for the time period of January 1982 through January 1986. SARP will develop generic bases to determine how safe the plants are and where and how their level of safety ought to be improved. The analysis to address these issues will be performed using improved probabilistic risk assessment methodology, as benchmarked to more exact data and analysis. There are thirteen program elements in the plan and the work is phased in two parts, with the first phase being completed in early 1984, at which time an assessment will be made whether or not any major changes will be recommended to the Commission for operating plants to handle severe accidents. Additionally at this time, all of the thirteen program elements in Chapter 5 will be reviewed and assessed in terms of how much additional work is necessary and where major impacts in probabilistic risk assessment might be achieved. Confirmatory research will be carried out in phase II to provide additional assurance on the appropriateness of phase I decisions. Most of this work will be concluded by early 1986

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

Energy Technology Data Exchange (ETDEWEB)

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

1981-01-15

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

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

International Nuclear Information System (INIS)

Fritsching, G.; Wolf, G.

1981-01-01

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

Evaluating Occupational Accidents and Their Indices In a Refining and Distributing Company of Petroleum Products of Mahshahr 2008-2010

Directory of Open Access Journals (Sweden)

A. Ansarimoghaddam

2016-01-01

Full Text Available Technological progress and oil industry development accompanies have a high rate of risk. This study was conducted with the aim of evaluating occupational accidents and related indicators for decreasing the number of damages by offering control measures. In this descriptive-analytical study, essential information was extracted from the records and agendas of the technical safety committee and the evaluation of accident frequency was done based on the accident type, the time and the location of its occurrence, environmental condition, root factors, and demographic variables of the injured. The relationship between repetition, severity, frequency of the accidents and marital status of the company’s personnel was also studied. Accident analysis was done by Chi square test . The occurrence of about 102 accidents was reported in an incident evaluation, which was done between the years 2008 and 2010. The average age of the injured was 29.1± 8.61. Accidents and clashes were 31.4% of the accidents and falling from height 21.6% of them. Many of the accidents occurred in plumbing activities (24.5%, tank construction (23.5%, and civil operations (15.7%. 47% of the accidents happened in 2009, 43% in 2010, and 10% in 2008. Occurrence rate of accidents was 48.1 and their intensity rate equaled 0.15 for one million working hours. The relationship of accident type and marital status was a meaningful relationship based on Chi-square test (P = 0.014; this test showed the relationship of accident type and its reasons a significant one as well. (P = 0.035 Considering the calculated coefficients and evaluated factors in this study, safety training, constant inquiring of sub-activities, inserting safety and HSE provisions and guidelines to the contractors’ contracts and monitoring their application would be effective in decreasing the number accidents. .

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.

Work safety in the light of technological progress in mining. [Poland, accident and technology assessment for period 1950 to 1972

Energy Technology Data Exchange (ETDEWEB)

Stecko, R.

1976-01-01

This paper discusses positive and negative aspects of technological progress in black coal mining in Poland in the period from 1950 to 1972 on the basis of statistical data supplied by various institutions. Twenty three indexes characterizing both work safety and technological progress in underground coal mines are used. Equations of linear and non-linear even regression and equations of linear and non-linear multiple regression are derived on the basis of statistical records of 23 years. The equations are used to estimate indexes of work safety on the basis of technical progress indexes. It is suggested that from among 8 indexes characterizing work safety 4 are most useful and important: danger index, absenteeism index, frequency of underground fire index, and frequency of rock burst index. Analysis of the regression equation shows that technological progress in underground coal mines in Poland has caused a decrease in indexes describing: frequency of work accidents, frequency of underground fires, frequency of rock burst, and danger of underground work; and an increase in indexes characterizing: seriousness of work accidents, and absenteeism caused by work accidents. (84 refs.)

Instrumentation availability during severe accidents for a boiling water reactor with a Mark I containment

International Nuclear Information System (INIS)

Arcieri, W.C.; Hanson, D.J.

1992-02-01

In support of the US Nuclear Regulatory Commission Accident Management Research Program, the availability of instruments to supply accident management information during a broad range of severe accidents is evaluated for a Boiling Water Reactor with a Mark I containment. Results from this evaluation include: (1) the identification of plant conditions that would impact instrument performance and information needs during severe accidents; (2) the definition of envelopes of parameters that would be important in assessing the performance of plant instrumentation for a broad range of severe accident sequences; and (3) assessment of the availability of plant instrumentation during severe accidents

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.

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.

Accident information needs

International Nuclear Information System (INIS)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-12-31

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

Accident information needs

Energy Technology Data Exchange (ETDEWEB)

Hanson, D.J.; Arcieri, W.C.; Ward, L.W.

1992-01-01

A Five-step methodology has been developed to evaluate information needs for nuclear power plants under accident conditions and the availability of plant instrumentation during severe accidents. Step 1 examines the credible accidents and their relationships to plant safety functions. Step 2 determines the information personnel involved in accident management will need to understand plant behavior. Step 3 determines the capability of the instrumentation to function properly under severe accident conditions. Step 4 determines the conditions expected during the identified severe accidents. Step 5 compares the instrument capabilities and the severe accident conditions to evaluate the availability of the instrumentation to supply needed plant information.

Multi-micronucleus cells related with viral diseases, detected in the study of children affected by the Chernobyl accident

International Nuclear Information System (INIS)

Garcia L, O.; Lamadrid, A.I.; Manzano, J.

1996-01-01

Cells with multiple chromosome aberrations have been observed in human peripheral blood lymphocytes. Different explanation have proposed, included hot particle induction in persons related to the Chernobyl accident. The frequency of chromosome aberration and micronuclei were established in 14 Ukrainian children with different hematological disorders. They arrived in Cuba thanks to the program by means of which medical attention is offered to children from areas affected by the Chernobyl accident. At least 500 metaphases and bi-nucleate cells were analyzed in each case. The detection of 4 cells with 7-11 micronuclei in a 14 year old boy with cat scratch disease was the most significant cytogenetical finding. The viral origin of the cat scratch disease has been reported, this suggested a viral etiology of the cells with multiple micronuclei. No rogue cells were detected. Cells with multiple micronuclei or rogue cells were not found in other patients from this group. (authors). 7 refs., 3 tabs

Association between knowledge and attitudes of school nurses towards epilepsy and the risk of accidents in Greek schools.

Science.gov (United States)

Toli, Theodora; Sourtzi, Panagiota; Tsoumakas, Konstantinos; Kalokerinou-Anagnostopoulou, Athena

2013-05-01

School nurses have the ability to enhance the knowledge and tolerance of an entire community and to form more positive and sensitized attitudes to future adults. The aim of this study was to investigate the relationship between the knowledge and attitudes of nurses and the frequency of accidents caused by epilepsy in Greek schools. Our sample consisted of 306 school nurses from all over the country. It was observed that the knowledge of school nurses on epilepsy was quite high, although there were specific aspects that raise concerns on their preparedness to respond to seizure-related emergencies, while their attitudes, although positive, still need improvement. Accidents caused by epilepsy were reported by half of the nurses, and prevention was considered of major importance. Therefore, organized continuous education programs and clear guidelines by the responsible authorities would help school nurses provide better services to students with epilepsy. Copyright © 2013 Elsevier Inc. All rights reserved.

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

Electrical equipment performance under severe accident conditions (BWR/Mark 1 plant analysis): Summary report

International Nuclear Information System (INIS)

Bennett, P.R.; Kolaczkowski, A.M.; Medford, G.T.

1986-09-01

The purpose of the Performance Evaluation of Electrical Equipment during Severe Accident States Program is to determine the performance of electrical equipment, important to safety, under severe accident conditions. In FY85, a method was devised to identify important electrical equipment and the severe accident environments in which the equipment was likely to fail. This method was used to evaluate the equipment and severe accident environments for Browns Ferry Unit 1, a BWR/Mark I. Following this work, a test plan was written in FY86 to experimentally determine the performance of one selected component to two severe accident environments

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