Reactivity insertion accident analysis

International Nuclear Information System (INIS)

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

1990-04-01

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

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

International Nuclear Information System (INIS)

Fonseca, Renato Alves da; Alvim, Antonio Carlos Marques

2005-01-01

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

The Development of Marine Accidents Human Reliability Assessment Approach: HEART Methodology and MOP Model

OpenAIRE

Ludfi Pratiwi Bowo; Wanginingastuti Mutmainnah; Masao Furusho

2017-01-01

Humans are one of the important factors in the assessment of accidents, particularly marine accidents. Hence, studies are conducted to assess the contribution of human factors in accidents. There are two generations of Human Reliability Assessment (HRA) that have been developed. Those methodologies are classified by the differences of viewpoints of problem-solving, as the first generation and second generation. The accident analysis can be determined using three techniques of analysis; sequen...

Safety analysis methodology for OPR 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun

2005-01-01

Full text: Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

Proposal of methodology of tsunami accident sequence analysis induced by earthquake using DQFM methodology

International Nuclear Information System (INIS)

Muta, Hitoshi; Muramatsu, Ken

2017-01-01

Since the Fukushima-Daiichi nuclear power station accident, the Japanese regulatory body has improved and upgraded the regulation of nuclear power plants, and continuous effort is required to enhance risk management in the mid- to long term. Earthquakes and tsunamis are considered as the most important risks, and the establishment of probabilistic risk assessment (PRA) methodologies for these events is a major issue of current PRA. The Nuclear Regulation Authority (NRA) addressed the PRA methodology for tsunamis induced by earthquakes, which is one of the methodologies that should be enhanced step by step for the improvement and maturity of PRA techniques. The AESJ standard for the procedure of seismic PRA for nuclear power plants in 2015 provides the basic concept of the methodology; however, details of the application to the actual plant PRA model have not been sufficiently provided. This study proposes a detailed PRA methodology for tsunamis induced by earthquakes using the DQFM methodology, which contributes to improving the safety of nuclear power plants. Furthermore, this study also states the issues which need more research. (author)

Development of a Methodology for VHTR Accident Consequence Assessment

Energy Technology Data Exchange (ETDEWEB)

Lee, Joeun; Kim, Jintae; Jae, Moosung [Hanyang University, Seoul (Korea, Republic of)

2016-05-15

The substitution of the VHTR for burning fossil fuels conserves these hydrocarbon resources for other uses and eliminates the emissions of greenhouse. In Korea, for these reasons, constructing the VHTR plan for hydrogen production is in progress. In this study, the consequence analysis for the off-site releases of radioactive materials during severe accidents has been performed using the level 3 PRA technology. The offsite consequence analysis for a VHTR using the MACCS code has been performed. Since the passive system such as the RCCS(Reactor Cavity Cooling System) are equipped, the frequency of occurrence of accidents has been evaluated to be very low. For further study, the assessment for characteristic of VHTR safety system and precise quantification of its accident scenarios is expected to conduct more certain consequence analysis. This methodology shown in this study might contribute to enhancing the safety of VHTR design by utilizing the results having far lower effect on the environment than the LWRs.

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

International Nuclear Information System (INIS)

Panayotov, Dobromir; Grief, Andrew; Merrill, Brad J.; Humrickhouse, Paul; Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon; Poitevin, Yves; Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard

2016-01-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

Methodology for accident analyses of fusion breeder blankets and its application to helium-cooled pebble bed blanket

Energy Technology Data Exchange (ETDEWEB)

Panayotov, Dobromir, E-mail: dobromir.panayotov@f4e.europa.eu [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Grief, Andrew [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Merrill, Brad J.; Humrickhouse, Paul [Idaho National Laboratory, PO Box 1625, Idaho Falls, ID (United States); Trow, Martin; Dillistone, Michael; Murgatroyd, Julian T.; Owen, Simon [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom); Poitevin, Yves [Fusion for Energy (F4E), Josep Pla, 2, Torres Diagonal Litoral B3, Barcelona E-08019 (Spain); Peers, Karen; Lyons, Alex; Heaton, Adam; Scott, Richard [Amec Foster Wheeler, Booths Park, Chelford Road, Knutsford WA16 8QZ, Cheshire (United Kingdom)

2016-11-01

Graphical abstract: - Highlights: • Test Blanket Systems (TBS) DEMO breeding blankets (BB) safety demonstration. • Comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena. • Development of accident analysis specifications (AAS) via the use of phenomena identification and ranking tables (PIRT). • PIRT application to identify required physical models for BB accidents analysis, code assessment and selection. • Development of MELCOR and RELAP5 codes TBS models. • Qualification of the models via comparison with finite element calculations, code-tocode comparisons, and sensitivity studies. - Abstract: ‘Fusion for Energy’ (F4E) is designing, developing, and implementing the European Helium-Cooled Lead-Lithium (HCLL) and Helium-Cooled Pebble-Bed (HCPB) Test Blanket Systems (TBSs) for ITER (Nuclear Facility INB-174). Safety demonstration is an essential element for the integration of these TBSs into ITER and accident analysis is one of its critical components. A systematic approach to accident analysis has been developed under the F4E contract on TBS safety analyses. F4E technical requirements, together with Amec Foster Wheeler and INL efforts, have resulted in a comprehensive methodology for fusion breeding blanket accident analysis that addresses the specificity of the breeding blanket designs, materials, and phenomena while remaining consistent with the approach already applied to ITER accident analyses. The methodology phases are illustrated in the paper by its application to the EU HCPB TBS using both MELCOR and RELAP5 codes.

Methodology to estimate the cost of the severe accidents risk / maximum benefit

International Nuclear Information System (INIS)

Mendoza, G.; Flores, R. M.; Vega, E.

2016-09-01

For programs and activities to manage aging effects, any changes to plant operations, inspections, maintenance activities, systems and administrative control procedures during the renewal period should be characterized, designed to manage the effects of aging as required by 10 Cfr Part 54 that could impact the environment. Environmental impacts significantly different from those described in the final environmental statement for the current operating license should be described in detail. When complying with the requirements of a license renewal application, the Severe Accident Mitigation Alternatives (SAMA) analysis is contained in a supplement to the environmental report of the plant that meets the requirements of 10 Cfr Part 51. In this paper, the methodology for estimating the cost of severe accidents risk is established and discussed, which is then used to identify and select the alternatives for severe accident mitigation, which are analyzed to estimate the maximum benefit that an alternative could achieve if this eliminate all risk. Using the regulatory analysis techniques of the US Nuclear Regulatory Commission (NRC) estimates the cost of severe accidents risk. The ultimate goal of implementing the methodology is to identify candidates for SAMA that have the potential to reduce the severe accidents risk and determine if the implementation of each candidate is cost-effective. (Author)

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

International Nuclear Information System (INIS)

Paula, H.M.

1984-01-01

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

Final report of the accident phenomenology and consequence (APAC) methodology evaluation. Spills Working Group

Energy Technology Data Exchange (ETDEWEB)

Brereton, S.; Shinn, J. [Lawrence Livermore National Lab., CA (United States); Hesse, D [Battelle Columbus Labs., OH (United States); Kaninich, D. [Westinghouse Savannah River Co., Aiken, SC (United States); Lazaro, M. [Argonne National Lab., IL (United States); Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

1997-08-01

The Spills Working Group was one of six working groups established under the Accident Phenomenology and Consequence (APAC) methodology evaluation program. The objectives of APAC were to assess methodologies available in the accident phenomenology and consequence analysis area and to evaluate their adequacy for use in preparing DOE facility safety basis documentation, such as Basis for Interim Operation (BIO), Justification for Continued Operation (JCO), Hazard Analysis Documents, and Safety Analysis Reports (SARs). Additional objectives of APAC were to identify development needs and to define standard practices to be followed in the analyses supporting facility safety basis documentation. The Spills Working Group focused on methodologies for estimating four types of spill source terms: liquid chemical spills and evaporation, pressurized liquid/gas releases, solid spills and resuspension/sublimation, and resuspension of particulate matter from liquid spills.

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1993-01-01

This paper will describe a methodology which has been developed to allow accident probabilities associated with one severity category scheme to be transferred to another severity category scheme, permitting some comparisons of different studies at the category level. In this methodology, the severity category schemes to be compared are mapped onto a common set of axes. The axes represent critical accident environments (e.g., impact, thermal, crush, puncture) and indicate the range of accident parameters from zero (no accident) to the most sever credible forces. The choice of critical accident environments for the axes depends on the package being transported and the mode of transportation. The accident probabilities associated with one scheme are then transferred to the other scheme. This transfer of category probabilities is based on the relationships of the critical accident parameters to probability of occurrence. The methodology can be employed to transfer any quantity between category schemes if the appropriate supporting information is available. (J.P.N.)

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)

APR1400 CEA Withdrawal at Power Accident Analysis using KNAP

International Nuclear Information System (INIS)

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

2006-01-01

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

Development of A Methodology for Assessing Various Accident Management Strategies Using Decision Tree Models

Energy Technology Data Exchange (ETDEWEB)

Kim, Nam Yeong; Kim, Jin Tae; Jae, Moo Sung [Hanyang University, Seoul (Korea, Republic of); Jerng, Dong Wook [Chung-Ang University, Seoul (Korea, Republic of)

2016-05-15

The purpose of ASP (Accident Sequence Precursor) analysis is to evaluate operational accidents in full power and low power operation by using PRA (Probabilistic Risk Assessment) technologies. The awareness of the importance of ASP analysis has been on rise. The methodology for ASP analysis has been developed in Korea, KINS (Korea Institute of Nuclear Safety) has managed KINS-ASP program since it was developed. In this study, we applied ASP analysis into operational accidents in full power and low power operation to quantify CCDP (Conditional Core Damage Probability). To reflect these 2 cases into PRA model, we modified fault trees and event trees of the existing PRA model. Also, we suggest the ASP regulatory system in the conclusion. In this study, we reviewed previous studies for ASP analysis. Based on it, we applied it into operational accidents in full power and low power operation. CCDP of these 2 cases are 1.195E-06 and 2.261E-03. Unlike other countries, there is no regulatory basis of ASP analysis in Korea. ASP analysis could detect the risk by assessing the existing operational accidents. ASP analysis can improve the safety of nuclear power plant by detecting, reviewing the operational accidents, and finally removing potential risk. Operator have to notify regulatory institute of operational accident before operator takes recovery work for the accident. After follow-up accident, they have to check precursors in data base to find similar accident.

A study on safety analysis methodology in spent fuel dry storage facility

Energy Technology Data Exchange (ETDEWEB)

Che, M. S.; Ryu, J. H.; Kang, K. M.; Cho, N. C.; Kim, M. S. [Hanyang Univ., Seoul (Korea, Republic of)

2004-02-15

Collection and review of the domestic and foreign technology related to spent fuel dry storage facility. Analysis of a reference system. Establishment of a framework for criticality safety analysis. Review of accident analysis methodology. Establishment of accident scenarios. Establishment of scenario analysis methodology.

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

International Nuclear Information System (INIS)

Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M.; Medford, G.T.

1990-06-01

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

Methodology used in IRSN nuclear accident cost estimates in France

International Nuclear Information System (INIS)

2015-01-01

This report describes the methodology used by IRSN to estimate the cost of potential nuclear accidents in France. It concerns possible accidents involving pressurized water reactors leading to radioactive releases in the environment. These accidents have been grouped in two accident families called: severe accidents and major accidents. Two model scenarios have been selected to represent each of these families. The report discusses the general methodology of nuclear accident cost estimation. The crucial point is that all cost should be considered: if not, the cost is underestimated which can lead to negative consequences for the value attributed to safety and for crisis preparation. As a result, the overall cost comprises many components: the most well-known is offsite radiological costs, but there are many others. The proposed estimates have thus required using a diversity of methods which are described in this report. Figures are presented at the end of this report. Among other things, they show that purely radiological costs only represent a non-dominant part of foreseeable economic consequences. (authors)

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1991-01-01

A methodology has been developed which allows the accident probabilities associated with one accident-severity category scheme to be transferred to another severity category scheme. The methodology requires that the schemes use a common set of parameters to define the categories. The transfer of accident probabilities is based on the relationships between probability of occurrence and each of the parameters used to define the categories. Because of the lack of historical data describing accident environments in engineering terms, these relationships may be difficult to obtain directly for some parameters. Numerical models or experienced judgement are often needed to obtain the relationships. These relationships, even if they are not exact, allow the accident probability associated with any severity category to be distributed within that category in a manner consistent with accident experience, which in turn will allow the accident probability to be appropriately transferred to a different category scheme

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).

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

Using MFM methodology to generate and define major accident scenarios for quantitative risk assessment studies

DEFF Research Database (Denmark)

Hua, Xinsheng; Wu, Zongzhi; Lind, Morten

2017-01-01

to calculate likelihood of each MAS. Combining the likelihood of each scenario with a qualitative risk matrix, each major accident scenario is thereby ranked for consideration for detailed consequence analysis. The methodology is successfully highlighted using part of BMA-process for production of hydrogen......Generating and defining Major Accident Scenarios (MAS) are commonly agreed as the key step for quantitative risk assessment (QRA). The aim of the study is to explore the feasibility of using Multilevel Flow Modeling (MFM) methodology to formulating MAS. Traditionally this is usually done based...

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

International Nuclear Information System (INIS)

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

2003-01-01

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

A methodology for analysing human errors of commission in accident scenarios for risk assessment

International Nuclear Information System (INIS)

Kim, J. H.; Jung, W. D.; Park, J. K

2003-01-01

As the concern on the impact of the operator's inappropriate interventions, so-called Errors Of Commissions(EOCs), on the plant safety has been raised, the interest in the identification and analysis of EOC events from the risk assessment perspective becomes increasing accordingly. To this purpose, we propose a new methodology for identifying and analysing human errors of commission that might be caused from the failures in situation assessment and decision making during accident progressions given an initiating event. The proposed methodology was applied to the accident scenarios of YGN 3 and 4 NPPs, which resulted in about 10 EOC situations that need careful attention

A systematic framework for effective uncertainty assessment of severe accident calculations; Hybrid qualitative and quantitative methodology

International Nuclear Information System (INIS)

Hoseyni, Seyed Mohsen; Pourgol-Mohammad, Mohammad; Tehranifard, Ali Abbaspour; Yousefpour, Faramarz

2014-01-01

This paper describes a systematic framework for characterizing important phenomena and quantifying the degree of contribution of each parameter to the output in severe accident uncertainty assessment. The proposed methodology comprises qualitative as well as quantitative phases. The qualitative part so called Modified PIRT, being a robust process of PIRT for more precise quantification of uncertainties, is a two step process for identifying and ranking based on uncertainty importance in severe accident phenomena. In this process identified severe accident phenomena are ranked according to their effect on the figure of merit and their level of knowledge. Analytical Hierarchical Process (AHP) serves here as a systematic approach for severe accident phenomena ranking. Formal uncertainty importance technique is used to estimate the degree of credibility of the severe accident model(s) used to represent the important phenomena. The methodology uses subjective justification by evaluating available information and data from experiments, and code predictions for this step. The quantitative part utilizes uncertainty importance measures for the quantification of the effect of each input parameter to the output uncertainty. A response surface fitting approach is proposed for estimating associated uncertainties with less calculation cost. The quantitative results are used to plan in reducing epistemic uncertainty in the output variable(s). The application of the proposed methodology is demonstrated for the ACRR MP-2 severe accident test facility. - Highlights: • A two stage framework for severe accident uncertainty analysis is proposed. • Modified PIRT qualitatively identifies and ranks uncertainty sources more precisely. • Uncertainty importance measure quantitatively calculates effect of each uncertainty source. • Methodology is applied successfully on ACRR MP-2 severe accident test facility

Methodology applied by IRSN for nuclear accident cost estimations in France

International Nuclear Information System (INIS)

2013-01-01

This report describes the methodology used by IRSN to estimate the cost of potential nuclear accidents in France. It concerns possible accidents involving pressurized water reactors leading to radioactive releases in the environment. These accidents have been grouped in two accident families called: severe accidents and major accidents. Two model scenarios have been selected to represent each of these families. The report discusses the general methodology of nuclear accident cost estimation. The crucial point is that all cost should be considered: if not, the cost is underestimated which can lead to negative consequences for the value attributed to safety and for crisis preparation. As a result, the overall cost comprises many components: the most well-known is offsite radiological costs, but there are many others. The proposed estimates have thus required using a diversity of methods which are described in this report. Figures are presented at the end of this report. Among other things, they show that purely radiological costs only represent a non-dominant part of foreseeable economic consequences

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

Methodological aspects to elaborate the management and procedure guides of severe accidents

International Nuclear Information System (INIS)

Gonzalez Gonzalez, F.; Jimenez Fernandez, A.

1995-01-01

The management guides in severe accidents are very important to know the procedures in these accidents. The present articles summarizes the methodological aspects to elaborate the management guides, in order to prevent the severe accidents

Methodology for flood risk analysis for nuclear power plants

International Nuclear Information System (INIS)

Wagner, D.P.; Casada, M.L.; Fussell, J.B.

1984-01-01

The methodology for flood risk analysis described here addresses the effects of a flood on nuclear power plant safety systems. Combining the results of this method with the probability of a flood allows the effects of flooding to be included in a probabilistic risk assessment. The five-step methodology includes accident sequence screening to focus the detailed analysis efforts on the accident sequences that are significantly affected by a flood event. The quantitative results include the flood's contribution to system failure probability, accident sequence occurrence frequency and consequence category occurrence frequency. The analysis can be added to existing risk assessments without a significant loss in efficiency. The results of two example applications show the usefulness of the methodology. Both examples rely on the Reactor Safety Study for the required risk assessment inputs and present changes in the Reactor Safety Study results as a function of flood probability

Accident selection methodology for TA-55 FSAR

International Nuclear Information System (INIS)

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

1995-01-01

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

Development of the methodology and approaches to validate safety and accident management

International Nuclear Information System (INIS)

Asmolov, V.G.

1997-01-01

The article compares the development of the methodology and approaches to validate the nuclear power plant safety and accident management in Russia and advanced industrial countries. It demonstrates that the development of methods of safety validation is dialectically related to the accumulation of the knowledge base on processes and events during NPP normal operation, transients and emergencies, including severe accidents. The article describes the Russian severe accident research program (1987-1996), the implementation of which allowed Russia to reach the world level of the safety validation efforts, presents future high-priority study areas. Problems related to possible approaches to the methodological accident management development are discussed. (orig.)

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

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

Level 2 PSA methodology and severe accident management

International Nuclear Information System (INIS)

1997-01-01

The objective of the work was to review current Level 2-PSA (Probabilistic Safety Assessment) methodologies and practices and to investigate how Level 2-PSA can support severe accident management programmes, i.e. the development, implementation, training and optimisation of accident management strategies and measures. For the most part, the presented material reflects the state in 1996. Current Level 2 PSA results and methodologies are reviewed and evaluated with respect to plant type specific and generic insights. Approaches and practices for using PSA results in the regulatory context and for supporting severe accident management programmes by input from level 2 PSAs are examined. The work is based on information contained in: PSA procedure guides, PSA review guides and regulatory guides for the use of PSA results in risk informed decision making; plant specific PSAs and PSA related literature exemplifying specific procedures, methods, analytical models, relevant input data and important results, use of computer codes and results of code calculations. The PSAs are evaluated with respect to results and insights. In the conclusion section, the present state of risk informed decision making, in particular in the level 2 domain, is described and substantiated by relevant examples

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.

Methodology and computational framework used for the US Department of Energy Environmental Restoration and Waste Management Programmatic Environmental Impact Statement accident analysis

International Nuclear Information System (INIS)

Mueller, C.; Roglans-Ribas, J.; Folga, S.; Huttenga, A.; Jackson, R.; TenBrook, W.; Russell, J.

1994-01-01

A methodology, computational framework, and integrated PC-based database have been developed to assess the risks of facility accidents in support of the US Department of Energy (DOE) Environmental Restoration and Waste Management Programmatic Environmental Impact Statement. The methodology includes the following interrelated elements: (1) screening of storage and treatment processes and related waste inventories to determine risk-dominant facilities across the DOE complex, (2) development and frequency estimation of the risk-dominant sequences of accidents, and (3) determination of the evolution of and final compositions of radiological or chemically hazardous source terms predicted to be released as a function of the storage inventory or treatment process throughput. The computational framework automates these elements to provide source term input for the second part of the analysis which includes (1) development or integration of existing site-specific demographics and meteorological data and calculation of attendant unit-risk factors and (2) assessment of the radiological or toxicological consequences of accident releases to the general public and to the occupational work force

Improved Methodology of MSLB M/E Release Analysis for OPR1000

International Nuclear Information System (INIS)

Park, Seok Jeong; Kim, Cheol Woo; Seo, Jong Tae

2006-01-01

A new mass and energy (M/E) release analysis methodology for the equipment environmental qualification (EEQ) on loss-of-coolant accident (LOCA) has been recently developed and adopted on small break LOCA EEQ. The new methodology for the M/E release analysis is extended to the M/E release analysis for the containment design for large break LOCA and the main steam line break (MSLB) accident, and named KIMERA (KOPEC Improved Mass and Energy Release Analysis) methodology. The computer code systems used in this methodology is RELAP5K/CONTEMPT4 (or RELAP5-ME) which couples RELAP5/MOD3.1/K with enhanced M/E model and LOCA long term model, and CONTEMPT4/ MOD5. This KIMERA methodology is applied to the MSLB M/E release analysis to evaluate the validation of KIMERA methodology for MSLB in containment design. The results are compared with the OPR 1000 FSAR

A methodology for analyzing precursors to earthquake-initiated and fire-initiated accident sequences

International Nuclear Information System (INIS)

Budnitz, R.J.; Lambert, H.E.; Apostolakis, G.

1998-04-01

This report covers work to develop a methodology for analyzing precursors to both earthquake-initiated and fire-initiated accidents at commercial nuclear power plants. Currently, the U.S. Nuclear Regulatory Commission sponsors a large ongoing project, the Accident Sequence Precursor project, to analyze the safety significance of other types of accident precursors, such as those arising from internally-initiated transients and pipe breaks, but earthquakes and fires are not within the current scope. The results of this project are that: (1) an overall step-by-step methodology has been developed for precursors to both fire-initiated and seismic-initiated potential accidents; (2) some stylized case-study examples are provided to demonstrate how the fully-developed methodology works in practice, and (3) a generic seismic-fragility date base for equipment is provided for use in seismic-precursors analyses. 44 refs., 23 figs., 16 tabs

A Methodology for Probabilistic Accident Management

International Nuclear Information System (INIS)

Munteanu, Ion; Aldemir, Tunc

2003-01-01

While techniques have been developed to tackle different tasks in accident management, there have been very few attempts to develop an on-line operator assistance tool for accident management and none that can be found in the literature that uses probabilistic arguments, which are important in today's licensing climate. The state/parameter estimation capability of the dynamic system doctor (DSD) approach is combined with the dynamic event-tree generation capability of the integrated safety assessment (ISA) methodology to address this issue. The DSD uses the cell-to-cell mapping technique for system representation that models the system evolution in terms of probability of transitions in time between sets of user-defined parameter/state variable magnitude intervals (cells) within a user-specified time interval (e.g., data sampling interval). The cell-to-cell transition probabilities are obtained from the given system model. The ISA follows the system dynamics in tree form and braches every time a setpoint for system/operator intervention is exceeded. The combined approach (a) can automatically account for uncertainties in the monitored system state, inputs, and modeling uncertainties through the appropriate choice of the cells, as well as providing a probabilistic measure to rank the likelihood of possible system states in view of these uncertainties; (b) allows flexibility in system representation; (c) yields the lower and upper bounds on the estimated values of state variables/parameters as well as their expected values; and (d) leads to fewer branchings in the dynamic event-tree generation. Using a simple but realistic pressurizer model, the potential use of the DSD-ISA methodology for on-line probabilistic accident management is illustrated

Risk analysis methodologies for the transportation of radioactive materials

International Nuclear Information System (INIS)

Geffen, C.A.

1983-05-01

Different methodologies have evolved for consideration of each of the many steps required in performing a transportation risk analysis. Although there are techniques that attempt to consider the entire scope of the analysis in depth, most applications of risk assessment to the transportation of nuclear fuel cycle materials develop specific methodologies for only one or two parts of the analysis. The remaining steps are simplified for the analyst by narrowing the scope of the effort (such as evaluating risks for only one material, or a particular set of accident scenarios, or movement over a specific route); performing a qualitative rather than a quantitative analysis (probabilities may be simply ranked as high, medium or low, for instance); or assuming some generic, conservative conditions for potential release fractions and consequences. This paper presents a discussion of the history and present state-of-the-art of transportation risk analysis methodologies. Many reports in this area were reviewed as background for this presentation. The literature review, while not exhaustive, did result in a complete representation of the major methods used today in transportation risk analysis. These methodologies primarily include the use of severity categories based on historical accident data, the analysis of specifically assumed accident sequences for the transportation activity of interest, and the use of fault or event tree analysis. Although the focus of this work has generally been on potential impacts to public groups, some effort has been expended in the estimation of risks to occupational groups in transportation activities

The development of a safety analysis methodology for the optimized power reactor 1000

International Nuclear Information System (INIS)

Hwang-Yong, Jun; Yo-Han, Kim

2005-01-01

Korea Electric Power Research Institute (KEPRI) has been developing inhouse safety analysis methodology based on the delicate codes available to KEPRI to overcome the problems arising from currently used vendor oriented methodologies. For the Loss of Coolant Accident (LOCA) analysis, the KREM (KEPRI Realistic Evaluation Methodology) has been developed based on the RELAP-5 code. The methodology was approved for the Westinghouse 3-loop plants by the Korean regulatory organization and the project to extent the methodology to the Optimized Power Reactor 1000 (OPR1000) has been ongoing since 2001. Also, for the Non-LOCA analysis, the KNAP (Korea Non-LOCA Analysis Package) has been developed using the UNICORN-TM code system. To demonstrate the feasibility of these codes systems and methodologies, some typical cases of the design basis accidents mentioned in the final safety analysis report (FSAR) were analyzed. (author)

Probabilistic Accident Progression Analysis with application to a LMFBR design

International Nuclear Information System (INIS)

Jamali, K.M.

1982-01-01

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

Estimating the potential impacts of a nuclear reactor accident: methodology and case studies

International Nuclear Information System (INIS)

Cartwright, J.V.; Beemiller, R.M.; Trott, E.A. Jr.; Younger, J.M.

1982-04-01

This monograph describes an industrial impact model that can be used to estimate the regional industry-specific impacts of disasters. Special attention is given to the impacts of possible nuclear reactor accidents. The monograph also presents three applications of the model. The impacts estimated in the case studies are based on (1) general information and reactor-specific data, supplied by the US Nuclear Regulatory Commission (NRC), (2) regional economic models derived from the Regional Input-Output Modeling System (RIMS II) developed at the Bureau of Economic Analysis (BEA), and (3) additional methodology developed especially for taking into account the unique characteristics of a nuclear reactor accident with respect to regional industrial activity

A Study on an Accident Diagnosis Methodology Using Influence Diagrams

International Nuclear Information System (INIS)

Kang, Kyungmin; Jae, Moosung

2006-01-01

For nuclear power plants, EOPs help operators to diagnose, control and mitigate accidents. However, it is very difficult that operators follow appropriate EOPs for accidents with similar symptoms in a given short period of time. Also EOPs are very complicated to follow and have many procedures to do. Therefore, if operators cannot diagnose correctly, the accident would become severe. Correct diagnostic action depends on the decision making ability of operators. Therefore, the methodology that can diagnose accidents quickly and help operators follow appropriate procedures should be developed. Due to the complexity of the tasks, it is very important to reduce human errors during diagnostic actions. In this study, to minimize human errors an accident diagnosis model has been constructed based on EOPs, accident symptoms and component reliabilities. For construction of model, Influence Diagrams have been applied. This decision-making tool consists of nodes and arcs. It is applicable to complicated situations, such as those required for developing strategies for managing severe accidents in nuclear power plants. And quantification of model has performed with total probability and Bayesian theorem. Through this quantification, the results should help operators diagnose complex situations

Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

International Nuclear Information System (INIS)

Liang, T.H.; Liang, K.S.; Cheng, C.K.; Pei, B.S.; Patelli, E.

2016-01-01

Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10 −3 . • The occurrence probability of the deterministic licensing sequence is 5.46 * 10 −5 . - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability of the

Risk-informed analysis of the large break loss of coolant accident and PCT margin evaluation with the RISMC methodology

Energy Technology Data Exchange (ETDEWEB)

Liang, T.H. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Liang, K.S., E-mail: ksliang@alum.mit.edu [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Cheng, C.K.; Pei, B.S. [Institute of Nuclear Engineering and Science, National Tsing Hua University, 101 Sec. 2, Kuang-Fu Road, Hsinchu 30013, Taiwan (China); Patelli, E. [Institute of Risk and Uncertainty, University of Liverpool, Room 610, Brodie Tower, L69 3GQ (United Kingdom)

2016-11-15

Highlights: • With RISMC methodology, both aleatory and epistemic uncertainties have been considered. • 14 probabilistically significant sequences have been identified and quantified. • A load spectrum for LBLOCA has been conducted with CPCT and SP of each dominant sequence. • Comparing to deterministic methodologies, the risk-informed PCT margin can be greater by 44–62 K. • The SP of the referred sequence to cover 99% in the load spectrum is only 5.07 * 10{sup −3}. • The occurrence probability of the deterministic licensing sequence is 5.46 * 10{sup −5}. - Abstract: For general design basis accidents, such as SBLOCA and LBLOCA, the traditional deterministic safety analysis methodologies are always applied to analyze events based on a so called surrogate or licensing sequence, without considering how low this sequence occurrence probability is. In the to-be-issued 10 CFR 50.46a, the LBLOCA will be categorized as accidents beyond design basis and the PCT margin shall be evaluated in a risk-informed manner. According to the risk-informed safety margin characterization (RISMC) methodology, a process has been suggested to evaluate the risk-informed PCT margin. Following the RISMC methodology, a load spectrum of PCT for LBLOCA has been generated for the Taiwan’s Maanshan Nuclear Power plant and 14 probabilistic significant sequences have been identified. It was observed in the load spectrum that the conditional PCT generally ascends with the descending sequence occurrence probability. With the load spectrum covering both aleatory and epistemic uncertainties, the risk-informed PCT margin can be evaluated by either expecting value estimation method or sequence probability coverage method. It was found that by comparing with the traditional deterministic methodology, the PCT margin evaluated by the RISMC methodology can be greater by 44–62 K. Besides, to have a cumulated occurrence probability over 99% in the load spectrum, the occurrence probability

Methodology to classify accident sequences of an Individual Plant Examination according to the severe releases for BWR type reactors

International Nuclear Information System (INIS)

Sandoval V, S.

2001-01-01

The Light Water Reactor (LWR) operation regulations require to every operating plant to perform of an Individual Plant Examination study (Ipe). One of the main purposes of an Ipe is t o gain a more quantitative understanding of the overall probabilities of core damage and fission product releases . Probabilistic Safety Analysis (PSA) methodologies and Severe Accident Analysis are used to perform Ipe studies. PSA methodologies are used to identify and analyse the set of event sequences that might originate the fission product release from a nuclear power plant; these methodologies are combinatorial in nature and generate thousands of sequences. Among other uses within an Ipe, severe accident simulations are used to determine the characteristics of the fission product release for the identified sequences and in this way, the releases can be understood and characterized. A vast amount of resources is required to simulate and analyse every Ipe sequence. This effort is unnecessary if similar sequences are grouped. The grouping scheme must achieve an efficient trade off between problem reduction and accuracy. The methodology presented in this work enables an accurate characterization and analysis of the Ipe fission product releases by using a reduced problem. The methodology encourages the use of specific plant simulations. (Author)

A methodology for the evaluation of fuel rod failures under transportation accidents

International Nuclear Information System (INIS)

Rashid, J.Y.R.; Machiels, A.J.

2004-01-01

Recent studies on long-term behavior of high-burnup spent fuel have shown that under normal conditions of stor-age, challenges to cladding integrity from various postulated damage mechanisms, such as delayed hydride crack-ing, stress-corrosion cracking and long-term creep, would not lead to any significant safety concerns during dry storage, and regulatory rules have subsequently been established to ensure that a compatible level of safety is maintained. However, similar safety assurances for spent fuel transportation have not yet been developed, and further studies are currently being conducted to evaluate the conditions under which transportation-related safety issues can be resolved. One of the issues presently under evaluation is the ability and the extent of the fuel as-semblies to maintain non-reconfigured geometry during transportation accidents. This evaluation may determine whether, or not, the shielding, confinement, and criticality safety evaluations can be performed assuming initial fuel assembly geometries. The degree to which spent fuel re-configuration could occur during a transportation accident would depend to a large degree on the number of fuel rod failures and the type and geometry of the failure modes. Such information can only be developed analytically, as there is no direct experimental data that can provide guidance on the level of damage that can be expected. To this end, the paper focuses on the development of a modeling and analysis methodology that deals with this general problem on a generic basis. First consideration is given to defining acci-dent loading that is equivalent to the bounding, although analytically intractable, hypothetical transportation acci-dent of a 9-meter drop onto essentially unyielding surface, which is effectively a condition for impact-limiters de-sign. Second, an analytically robust material constitutive model, an essential element in a successful structural analysis, is required. A material behavior model

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.

Development of Advanced Non-LOCA Analysis Methodology for Licensing

International Nuclear Information System (INIS)

Jang, Chansu; Um, Kilsup; Choi, Jaedon

2008-01-01

KNF is developing a new design methodology on the Non-LOCA analysis for the licensing purpose. The code chosen is the best-estimate transient analysis code RETRAN and the OPR1000 is aimed as a target plant. For this purpose, KNF prepared a simple nodal scheme appropriate to the licensing analyses and developed the designer-friendly analysis tool ASSIST (Automatic Steady-State Initialization and Safety analysis Tool). To check the validity of the newly developed methodology, the single CEA withdrawal and the locked rotor accidents are analyzed by using a new methodology and are compared with current design results. Comparison results show a good agreement and it is concluded that the new design methodology can be applied to the licensing calculations for OPR1000 Non-LOCA

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Development of Accident Scenarios and Quantification Methodology for RAON Accelerator

International Nuclear Information System (INIS)

Lee, Yongjin; Jae, Moosung

2014-01-01

The RIsp (Rare Isotope Science Project) plans to provide neutron-rich isotopes (RIs) and stable heavy ion beams. The accelerator is defined as radiation production system according to Nuclear Safety Law. Therefore, it needs strict operate procedures and safety assurance to prevent radiation exposure. In order to satisfy this condition, there is a need for evaluating potential risk of accelerator from the design stage itself. Though some of PSA researches have been conducted for accelerator, most of them focus on not general accident sequence but simple explanation of accident. In this paper, general accident scenarios are developed by Event Tree and deduce new quantification methodology of Event Tree. In this study, some initial events, which may occur in the accelerator, are selected. Using selected initial events, the accident scenarios of accelerator facility are developed with Event Tree. These results can be used as basic data of the accelerator for future risk assessments. After analyzing the probability of each heading, it is possible to conduct quantification and evaluate the significance of the accident result. If there is a development of the accident scenario for external events, risk assessment of entire accelerator facility will be completed. To reduce the uncertainty of the Event Tree, it is possible to produce a reliable data via the presented quantification techniques

Simplified methodology for Angra 1 containment analysis

International Nuclear Information System (INIS)

Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

1991-08-01

A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminary evaluation of the Angra 1 global parameters. (author)

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

International Nuclear Information System (INIS)

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

2008-01-01

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

Upgrading the safety toolkit: Initiatives of the accident analysis subgroup

International Nuclear Information System (INIS)

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

1999-01-01

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

Preliminary safety analysis methodology for the SMART

Energy Technology Data Exchange (ETDEWEB)

Bae, Kyoo Hwan; Chung, Y. J.; Kim, H. C.; Sim, S. K.; Lee, W. J.; Chung, B. D.; Song, J. H. [Korea Atomic Energy Research Institute, Taejeon (Korea)

2000-03-01

This technical report was prepared for a preliminary safety analysis methodology of the 330MWt SMART (System-integrated Modular Advanced ReacTor) which has been developed by Korea Atomic Energy Research Institute (KAERI) and funded by the Ministry of Science and Technology (MOST) since July 1996. This preliminary safety analysis methodology has been used to identify an envelope for the safety of the SMART conceptual design. As the SMART design evolves, further validated final safety analysis methodology will be developed. Current licensing safety analysis methodology of the Westinghouse and KSNPP PWRs operating and under development in Korea as well as the Russian licensing safety analysis methodology for the integral reactors have been reviewed and compared to develop the preliminary SMART safety analysis methodology. SMART design characteristics and safety systems have been reviewed against licensing practices of the PWRs operating or KNGR (Korean Next Generation Reactor) under construction in Korea. Detailed safety analysis methodology has been developed for the potential SMART limiting events of main steam line break, main feedwater pipe break, loss of reactor coolant flow, CEA withdrawal, primary to secondary pipe break and the small break loss of coolant accident. SMART preliminary safety analysis methodology will be further developed and validated in parallel with the safety analysis codes as the SMART design further evolves. Validated safety analysis methodology will be submitted to MOST as a Topical Report for a review of the SMART licensing safety analysis methodology. Thus, it is recommended for the nuclear regulatory authority to establish regulatory guides and criteria for the integral reactor. 22 refs., 18 figs., 16 tabs. (Author)

[Methodologies for estimating the indirect costs of traffic accidents].

Science.gov (United States)

Carozzi, Soledad; Elorza, María Eugenia; Moscoso, Nebel Silvana; Ripari, Nadia Vanina

2017-01-01

Traffic accidents generate multiple costs to society, including those associated with the loss of productivity. However, there is no consensus about the most appropriate methodology for estimating those costs. The aim of this study was to review methods for estimating indirect costs applied in crash cost studies. A thematic review of the literature was carried out between 1995 and 2012 in PubMed with the terms cost of illness, indirect cost, road traffic injuries, productivity loss. For the assessment of costs we used the the human capital method, on the basis of the wage-income lost during the time of treatment and recovery of patients and caregivers. In the case of premature death or total disability, the discount rate was applied to obtain the present value of lost future earnings. The computed years arose by subtracting to life expectancy at birth the average age of those affected who are not incorporated into the economically active life. The interest in minimizing the problem is reflected in the evolution of the implemented methodologies. We expect that this review is useful to estimate efficiently the real indirect costs of traffic accidents.

The analysis of RWAP(Rod Withdrawal at Power) using the KEPRI methodology

International Nuclear Information System (INIS)

Yang, C. K.; Kim, Y. H.

2001-01-01

KEPRI developed new methodology which was based on RASP(Reactor Analysis Support Package). In this paper, The analysis of RWAP(Rod Withdrawal at Power) accident which can result in reactivity and power distribution anomaly was performed using the KEPRI methodology. The calculation describes RWAP transient and documents the analysis, including the computer code modeling assumptions and input parameters used in the analysis. To validity for the new methodology, the result of calculation was compared with FSAR. As compared with FSAR, result of the calculation using the KEPRI Methodology is similar to FSAR's. And result of the sensitivity of postulated parameters were similar to the existing methodology

Methodology for the Assessment of Confidence in Safety Margin for Small Break Loss of Coolant Accident Sequences

Energy Technology Data Exchange (ETDEWEB)

Nagrale, D. B.; Prasad, M.; Rao, R. S.; Gaikwad, A.J., E-mail: avinashg@aerb.gov.in [Nuclear Safety Analysis Division, Atomic Energy Regulatory Board, Mumbai (India)

2014-10-15

Deterministic Safety Analysis and Probabilistic Safety Assessment (PSA) analyses are used concurrently to assess the Nuclear Power Plant (NPP) safety. The conventional deterministic analysis is conservative. The best estimate plus uncertainty analysis is increasingly being used for deterministic calculation in NPPs. The PSA methodology aims to be as realistic as possible while integrating information about accident phenomena, plant design, operating practices, component reliability and human behaviour. The peak clad temperature (PCT) distribution provides an insight into the confidence in safety margin for an initiating event. The paper deals with the concept of calculating the peak clad temperature with 95 percent confidence and 95 percent probability (PCT{sub 95/95}) in small break loss of coolant accident (SBLOCA) and methodologies for assessing safety margin. Five input parameters mainly, nominal power level, decay power, fuel clad gap conductivity, fuel thermal conductivity and discharge coefficient, were selected. A Uniform probability density function was assigned to the uncertain parameters and these uncertainties are propagated using Latin Hypercube Sampling (LHS) technique. The sampled data for 5 parameters were randomly mixed by LHS to obtain 25 input sets. A non-core damage accident sequence was selected from the SBLOCA event tree of a typical VVER study to estimate the PCTs and safety margin. A Kolmogorov– Smirnov goodness-of-fit test was carried out for PCTs. The smallest value of safety margin would indicate the robustness of the system with 95% confidence and 95% probability. Regression analysis was also carried out using 1000 sample size for the estimating PCTs. Mean, variance and finally safety margin were analysed. (author)

Environmental impact statement analysis: dose methodology

International Nuclear Information System (INIS)

Mueller, M.A.; Strenge, D.L.; Napier, B.A.

1981-01-01

Standardized sections and methodologies are being developed for use in environmental impact statements (EIS) for activities to be conducted on the Hanford Reservation. Five areas for standardization have been identified: routine operations dose methodologies, accident dose methodology, Hanford Site description, health effects methodology, and socioeconomic environment for Hanford waste management activities

Severe accident analysis for level 2 PSA of SMART reactor

Energy Technology Data Exchange (ETDEWEB)

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

2010-12-15

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

Simplified methodology for analysis of Angra-1 containing

International Nuclear Information System (INIS)

Neves Conti, T. das; Souza, A.L. de; Sabundjian, G.

1988-01-01

A simplified methodology of analysis was developed to simulate a Large Break Loss of Coolant Accident in the Angra 1 Nuclear Power Station. Using the RELAP5/MOD1, RELAP4/MOD5 and CONTEMPT-LT Codes, the time the variation of pressure and temperature in the containment was analysed. The obtained data was compared with the Angra 1 Final Safety Analysis Report, and too those calculated by a Detailed Model. The results obtained by this new methodology such as the small computational time of simulation, were satisfactory when getting the preliminar avaliation of the Angra 1 global parameters. (author) [pt

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

International Nuclear Information System (INIS)

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

2001-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2001-07-01

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

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

International Nuclear Information System (INIS)

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

2014-01-01

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

Analysis of Aircraft Crash Accident for WETF

International Nuclear Information System (INIS)

Jordan, Hans

2001-01-01

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

Methodologies for uncertainty analysis in the level 2 PSA and their implementation procedures

International Nuclear Information System (INIS)

Ahn, Kwang Il; Yang, Joon Eun; Kim, Dong Ha

2002-04-01

Main purpose of this report to present standardized methodologies for uncertainty analysis in the Level 2 Probabilistic Safety Assessment (PSA) and their implementation procedures, based on results obtained through a critical review of the existing methodologies for the analysis of uncertainties employed in the Level 2 PSA, especially Accident Progression Event Tree (APET). Uncertainties employed in the Level 2 PSA, quantitative expressions of overall knowledge of analysts' and experts' participating in the probabilistic quantification process of phenomenological accident progressions ranging from core melt to containment failure, their numerical values are directly related to the degree of confidence that the analyst has that a given phenomenological event or accident process will or will not occur, or analyst's subjective probabilities of occurrence. These results that are obtained from Level 2 PSA uncertainty analysis, become an essential contributor to the plant risk, in addition to the Level 1 PSA and Level 3 PSA uncertainties. Uncertainty analysis methodologies and their implementation procedures presented in this report was prepared based on the following criteria: 'uncertainty quantification process must be logical, scrutable, complete, consistent and in an appropriate level of detail, as mandated by the Level 2 PSA objectives'. For the aforementioned purpose, this report deals mainly with (1) summary of general or Level 2 PSA specific uncertainty analysis methodologies, (2) selection of phenomenological branch events for uncertainty analysis in the APET, methodology for quantification of APET uncertainty inputs and its implementation procedure, (3) statistical propagation of uncertainty inputs through APET and its implementation procedure, and (4) formal procedure for quantification of APET uncertainties and source term categories (STCs) through the Level 2 PSA quantification codes

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

International Nuclear Information System (INIS)

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

2003-01-01

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

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

International Nuclear Information System (INIS)

1975-10-01

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

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

International Nuclear Information System (INIS)

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

1996-01-01

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

Analysis of the falling accident of a Control bar to power with adiabatic and realistic methodologies; Analisis del Accidente de Caida de una Barra de Control a Potencia con Metodologias Adiabatica y Realista

Energy Technology Data Exchange (ETDEWEB)

Paredes, C.; Trueba, M.

2014-07-01

The CRDA accident in BWR plants was analysed historically in situations of stop cold and stop hot with adiabatic methods, to have considered these situations the more limiting for this type of event. The proposed analysis is done with the two methodologies in order to show that this accident is effectively limiting how much lower is the power, as well as the conservatism of the adiabatic approximation. (Author)

Applying the AcciMap methodology to investigate the tragic Sewol Ferry accident in South Korea.

Science.gov (United States)

Lee, Samuel; Moh, Young Bo; Tabibzadeh, Maryam; Meshkati, Najmedin

2017-03-01

This study applies the AcciMap methodology, which was originally proposed by Professor Jens Rasmussen (1997), to the analysis of the tragic Sewol Ferry accident in South Korea on April 16, 2014, which killed 304 mostly young people and is considered as a national disaster in that country. This graphical representation, by incorporating associated socio-technical factors into an integrated framework, provides a big-picture to illustrate the context in which an accident occurred as well as the interactions between different levels of the studied system that resulted in that event. In general, analysis of past accidents within the stated framework can define the patterns of hazards within an industrial sector. Such analysis can lead to the definition of preconditions for safe operations, which is a main focus of proactive risk management systems. In the case of the Sewol Ferry accident, a lot of the blame has been placed on the Sewol's captain and its crewmembers. However, according to this study, which relied on analyzing all available sources published in English and Korean, the disaster is the result of a series of lapses and disregards for safety across different levels of government and regulatory bodies, Chonghaejin Company, and the Sewol's crewmembers. The primary layers of the AcciMap framework, which include the political environment and non-proactive governmental body; inadequate regulations and their lax oversight and enforcement; poor safety culture; inconsideration of human factors issues; and lack of and/or outdated standard operating and emergency procedures were not only limited to the maritime industry in South Korea, and the Sewol Ferry accident, but they could also subject any safety-sensitive industry anywhere in the world. Copyright © 2016 Elsevier Ltd. All rights reserved.

Methodology to estimate the cost of the severe accidents risk / maximum benefit; Metodologia para estimar el costo del riesgo de accidentes severos / beneficio maximo

Energy Technology Data Exchange (ETDEWEB)

Mendoza, G.; Flores, R. M.; Vega, E., E-mail: gozalo.mendoza@inin.gob.mx [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

For programs and activities to manage aging effects, any changes to plant operations, inspections, maintenance activities, systems and administrative control procedures during the renewal period should be characterized, designed to manage the effects of aging as required by 10 Cfr Part 54 that could impact the environment. Environmental impacts significantly different from those described in the final environmental statement for the current operating license should be described in detail. When complying with the requirements of a license renewal application, the Severe Accident Mitigation Alternatives (SAMA) analysis is contained in a supplement to the environmental report of the plant that meets the requirements of 10 Cfr Part 51. In this paper, the methodology for estimating the cost of severe accidents risk is established and discussed, which is then used to identify and select the alternatives for severe accident mitigation, which are analyzed to estimate the maximum benefit that an alternative could achieve if this eliminate all risk. Using the regulatory analysis techniques of the US Nuclear Regulatory Commission (NRC) estimates the cost of severe accidents risk. The ultimate goal of implementing the methodology is to identify candidates for SAMA that have the potential to reduce the severe accidents risk and determine if the implementation of each candidate is cost-effective. (Author)

Quantifying reactor safety margins: Application of code scaling, applicability, and uncertainty evaluation methodology to a large-break, loss-of-coolant accident

International Nuclear Information System (INIS)

Boyack, B.; Duffey, R.; Wilson, G.; Griffith, P.; Lellouche, G.; Levy, S.; Rohatgi, U.; Wulff, W.; Zuber, N.

1989-12-01

The US Nuclear Regulatory Commission (NRC) has issued a revised rule for loss-of-coolant accident/emergency core cooling system (ECCS) analysis of light water reactors to allow the use of best-estimate computer codes in safety analysis as an option. A key feature of this option requires the licensee to quantify the uncertainty of the calculations and include that uncertainty when comparing the calculated results with acceptance limits provided in 10 CFR Part 50. To support the revised ECCS rule and illustrate its application, the NRC and its contractors and consultants have developed and demonstrated an uncertainty evaluation methodology called code scaling, applicability, and uncertainty (CSAU). The CSAU methodology and an example application described in this report demonstrate that uncertainties in complex phenomena can be quantified. The methodology is structured, traceable, and practical, as is needed in the regulatory arena. The methodology is systematic and comprehensive as it addresses and integrates the scenario, experiments, code, and plant to resolve questions concerned with: (a) code capability to scale-up processes from test facility to full-scale nuclear power plants; (b) code applicability to safety studies of a postulated accident scenario in a specified nuclear power plant; and (c) quantifying uncertainties of calculated results. 127 refs., 55 figs., 40 tabs

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

International Nuclear Information System (INIS)

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

2009-01-01

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

Accident analysis for nuclear power plants

International Nuclear Information System (INIS)

2002-01-01

Deterministic safety analysis (frequently referred to as accident analysis) is an important tool for confirming the adequacy and efficiency of provisions within the defence in depth concept for the safety of nuclear power plants (NPPs). Owing to the close interrelation between accident analysis and safety, an analysis that lacks consistency, is incomplete or is of poor quality is considered a safety issue for a given NPP. Developing IAEA guidance documents for accident analysis is thus an important step towards resolving this issue. Requirements and guidelines pertaining to the scope and content of accident analysis have, in the past, been partially described in various IAEA documents. Several guidelines relevant to WWER and RBMK type reactors have been developed within the IAEA Extrabudgetary Programme on the Safety of WWER and RBMK NPPs. To a certain extent, accident analysis is also covered in several documents of the revised NUSS series, for example, in the Safety Requirements on Safety of Nuclear Power Plants: Design (NS-R-1) and in the Safety Guide on Safety Assessment and Verification for Nuclear Power Plants (NS-G-1.2). Consistent with these documents, the IAEA has developed the present Safety Report on Accident Analysis for Nuclear Power Plants. Many experts have contributed to the development of this Safety Report. Besides several consultants meetings, comments were collected from more than fifty selected organizations. The report was also reviewed at the IAEA Technical Committee Meeting on Accident Analysis held in Vienna from 30 August to 3 September 1999. The present IAEA Safety Report is aimed at providing practical guidance for performing accident analyses. The guidance is based on present good practice worldwide. The report covers all the steps required to perform accident analyses, i.e. selection of initiating events and acceptance criteria, selection of computer codes and modelling assumptions, preparation of input data and presentation of the

Application of human reliability analysis methodology of second generation

International Nuclear Information System (INIS)

Ruiz S, T. de J.; Nelson E, P. F.

2009-10-01

The human reliability analysis (HRA) is a very important part of probabilistic safety analysis. The main contribution of HRA in nuclear power plants is the identification and characterization of the issues that are brought together for an error occurring in the human tasks that occur under normal operation conditions and those made after abnormal event. Additionally, the analysis of various accidents in history, it was found that the human component has been a contributing factor in the cause. Because of need to understand the forms and probability of human error in the 60 decade begins with the collection of generic data that result in the development of the first generation of HRA methodologies. Subsequently develop methods to include in their models additional performance shaping factors and the interaction between them. So by the 90 mid, comes what is considered the second generation methodologies. Among these is the methodology A Technique for Human Event Analysis (ATHEANA). The application of this method in a generic human failure event, it is interesting because it includes in its modeling commission error, the additional deviations quantification to nominal scenario considered in the accident sequence of probabilistic safety analysis and, for this event the dependency actions evaluation. That is, the generic human failure event was required first independent evaluation of the two related human failure events . So the gathering of the new human error probabilities involves the nominal scenario quantification and cases of significant deviations considered by the potential impact on analyzed human failure events. Like probabilistic safety analysis, with the analysis of the sequences were extracted factors more specific with the highest contribution in the human error probabilities. (Author)

A DOE-STD-3009 hazard and accident analysis methodology for non-reactor nuclear facilities

International Nuclear Information System (INIS)

MAHN, JEFFREY A.; WALKER, SHARON ANN

2000-01-01

This paper demonstrates the use of appropriate consequence evaluation criteria in conjunction with generic likelihood of occurrence data to produce consistent hazard analysis results for nonreactor nuclear facility Safety Analysis Reports (SAR). An additional objective is to demonstrate the use of generic likelihood of occurrence data as a means for deriving defendable accident sequence frequencies, thereby enabling the screening of potentially incredible events ( -6 per year) from the design basis accident envelope. Generic likelihood of occurrence data has been used successfully in performing SAR hazard and accident analyses for two nonreactor nuclear facilities at Sandia National Laboratories. DOE-STD-3009-94 addresses and even encourages use of a qualitative binning technique for deriving and ranking nonreactor nuclear facility risks. However, qualitative techniques invariably lead to reviewer requests for more details associated with consequence or likelihood of occurrence bin assignments in the test of the SAR. Hazard analysis data displayed in simple worksheet format generally elicits questions about not only the assumptions behind the data, but also the quantitative bases for the assumptions themselves (engineering judgment may not be considered sufficient by some reviewers). This is especially true where the criteria for qualitative binning of likelihood of occurrence involves numerical ranges. Oftentimes reviewers want to see calculations or at least a discussion of event frequencies or failure probabilities to support likelihood of occurrence bin assignments. This may become a significant point of contention for events that have been binned as incredible. This paper will show how the use of readily available generic data can avoid many of the reviewer questions that will inevitably arise from strictly qualitative analyses, while not significantly increasing the overall burden on the analyst

Prediction of road accidents: A Bayesian hierarchical approach

DEFF Research Database (Denmark)

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

2013-01-01

the expected number of accidents in which an injury has occurred and the expected number of light, severe and fatally injured road users. Additionally, the methodology is used for geo-referenced identification of road sections with increased occurrence probabilities of injury accident events on a road link......In this paper a novel methodology for the prediction of the occurrence of road accidents is presented. The methodology utilizes a combination of three statistical methods: (1) gamma-updating of the occurrence rates of injury accidents and injured road users, (2) hierarchical multivariate Poisson......-lognormal regression analysis taking into account correlations amongst multiple dependent model response variables and effects of discrete accident count data e.g. over-dispersion, and (3) Bayesian inference algorithms, which are applied by means of data mining techniques supported by Bayesian Probabilistic Networks...

Advantages of Westinghouse BWR control rod drop accidents methodology utilizing integrated POLCA-T code

International Nuclear Information System (INIS)

Panayotov, Dobromir

2008-01-01

The paper focuses on the activities pursued by Westinghouse in the development and licensing of POLCA-T code Control Rod Drop Accident (CRDA) Methodology. The comprehensive CRDA methodology that utilizes PHOENIX4/POLCA7/POLCA-T calculation chain foresees complete cycle-specific analysis. The methodology consists of determination of candidates of control rods (CR) that could cause a significant reactivity excursion if dropped throughout the entire fuel cycle, selection of limiting initial conditions for CRDA transient simulation and transient simulation itself. The Westinghouse methodology utilizes state-of-the-art methods. Unnecessary conservatisms in the methodology have been avoided to allow the accurate prediction of margin to design bases. This is mainly achieved by using the POLCA-T code for dynamic CRDA evaluations. The code belongs to the same calculation chain that is used for core design. Thus the very same reactor, core, cycle and fuel data base is used. This allows also reducing the uncertainties of input data and parameters that determine the energy deposition in the fuel. Uncertainty treatment, very selective use of conservatisms, selection of the initial conditions for limiting case analyses, incorporation into POLCA-T code models of the licensed fuel performance code are also among the means of performing realistic CRDA transient analyses. (author)

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

International Nuclear Information System (INIS)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations

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)

Systematic approach for assessment of accident risks in chemical and nuclear processing

International Nuclear Information System (INIS)

Senne Junior, Murillo

2003-07-01

The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of systematizing

A human reliability analysis of the Three Mile power plant accident considering the THERP and ATHEANA methodologies

International Nuclear Information System (INIS)

Fonseca, Renato Alves da

2004-03-01

The main purpose of this work is the study of human reliability using the THERP (Technique for Human Error Prediction) and ATHEANA methods (A Technique for Human Error Analysis), and some tables and also, from case studies presented on the THERP Handbook to develop a qualitative and quantitative study of nuclear power plant accident. This accident occurred in the TMI (Three Mile Island Unit 2) power plant, PWR type plant, on March 28th, 1979. The accident analysis has revealed a series of incorrect actions, which resulted in the Unit 2 shut down and permanent loss of the reactor. This study also aims at enhancing the understanding of the THERP method and ATHEANA, and of its practical applications. In addition, it is possible to understand the influence of plant operational status on human failures and of these on equipment of a system, in this case, a nuclear power plant. (author)

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

International Nuclear Information System (INIS)

Zhang Shishun; Zhou Zhou; Xiao Min

2004-01-01

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

A Study for Appropriateness of National Nuclear Policy by using Economic Analysis Methodology after Fukushima accident

International Nuclear Information System (INIS)

Shim, Jong Myoung; Roh, Myung Sub

2013-01-01

The aim of this paper is to clarify the appropriateness of national nuclear policy in BPE of Korea from an economic perspective. To do this, this paper only focus on the economic analysis methodology without any considering other conditions such as political, cultural, or historical things. In a number of countries, especially Korea, nuclear energy policy is keeping the status quo after Fukushima accident. However the nation's nuclear policy may vary depending on the choice of people. Thus, to make the right decisions, it is important to deliver accurate information and knowledge about nuclear energy to the people. As proven in this paper, the levelized cost of nuclear power is the most inexpensive among the base load units. As the reliance on nuclear power is getting stronger through the economic logic, the nuclear safety and environmental elements will be strengthened. Based on this, national nuclear policy should be promoted. In the aftermath of the Fukushima accident recognized as the world's worst nuclear disaster since the Chernobyl, there are some changes in the nuclear energy policy of various countries. Germany, for example, called a halt to operate Nuclear Power Plant (NPP) which accounts for about 7.5% of the national power generation capacity of 6.3GW. In developing countries such as China and India they conducted the safety check of the nuclear power plants again before preceding their nuclear business. Korea government announced 'The 6th Basic Plan for Long-term Electricity Supply and Demand (BPE)', considering the safety and general public acceptance of the nuclear power plants. According to BPE, they postponed a plan for additional NPP construction, except for constructions that had been already reflected in the 5th BPE. All told, the responses for nuclear energy policy of countries are different depending on their own circumstances

A Study for Appropriateness of National Nuclear Policy by using Economic Analysis Methodology after Fukushima accident

Energy Technology Data Exchange (ETDEWEB)

Shim, Jong Myoung; Roh, Myung Sub [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2013-10-15

The aim of this paper is to clarify the appropriateness of national nuclear policy in BPE of Korea from an economic perspective. To do this, this paper only focus on the economic analysis methodology without any considering other conditions such as political, cultural, or historical things. In a number of countries, especially Korea, nuclear energy policy is keeping the status quo after Fukushima accident. However the nation's nuclear policy may vary depending on the choice of people. Thus, to make the right decisions, it is important to deliver accurate information and knowledge about nuclear energy to the people. As proven in this paper, the levelized cost of nuclear power is the most inexpensive among the base load units. As the reliance on nuclear power is getting stronger through the economic logic, the nuclear safety and environmental elements will be strengthened. Based on this, national nuclear policy should be promoted. In the aftermath of the Fukushima accident recognized as the world's worst nuclear disaster since the Chernobyl, there are some changes in the nuclear energy policy of various countries. Germany, for example, called a halt to operate Nuclear Power Plant (NPP) which accounts for about 7.5% of the national power generation capacity of 6.3GW. In developing countries such as China and India they conducted the safety check of the nuclear power plants again before preceding their nuclear business. Korea government announced 'The 6th Basic Plan for Long-term Electricity Supply and Demand (BPE)', considering the safety and general public acceptance of the nuclear power plants. According to BPE, they postponed a plan for additional NPP construction, except for constructions that had been already reflected in the 5th BPE. All told, the responses for nuclear energy policy of countries are different depending on their own circumstances.

Source term and radiological consequence evaluation for nuclear accidents using a 'hand type' methodology

International Nuclear Information System (INIS)

Margeanu, Sorin; Tatiana, Angelescu

2005-01-01

In the last decades, hand type calculations have been replaced by computerized solutions, which are much more accurate, but the preparation of an input to run the code can be a time consuming process and can require a laborious work. This is why, a place for hand calculation based on nomograms still exist in some areas. An example is emergency response to an accidental release of radioactive contaminants when the health of persons close to the accident site might be at risk. In this case, results from computerized accident consequences assessment models may be delayed due to the equipment malfunction or the time required developing minimal input files and performing the calculations (typically more than five minutes). A simple nomogram (developed using computerized dispersion model calculations) can provide dispersion and dose estimates within a minute. The paper presents the methodology used for these 'hand type' calculation and the nomograms, figures and tables used to evaluate the dose to an individual close to the release point. In order to illustrate the use of methodology, a hypothetical severe accident scenario involving 14-MW INR-TRIGA research reactor was considered. (authors)

Accident analysis. A review of the various accidents classifications

International Nuclear Information System (INIS)

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

1982-01-01

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

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

International Nuclear Information System (INIS)

Dubey, S.K.; Petruzzi, A.; Giannotti, W.; D'Auria, F.

2006-01-01

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

Phenomenological analyses and their application to the Defense Waste Processing Facility probabilistic safety analysis accident progression event tree. Revision 1

International Nuclear Information System (INIS)

Kalinich, D.A.; Thomas, J.K.; Gough, S.T.; Bailey, R.T.; Kearnaghan, D.P.

1995-01-01

In the Defense Waste Processing Facility (DWPF) Safety Analysis Reports (SARs) for the Savannah River Site (SRS), risk-based perspectives have been included per US Department of Energy (DOE) Order 5480.23. The NUREG-1150 Level 2/3 Probabilistic Risk Assessment (PRA) methodology was selected as the basis for calculating facility risk. The backbone of this methodology is the generation of an Accident Progression Event Tree (APET), which is solved using the EVNTRE computer code. To support the development of the DWPF APET, deterministic modeling of accident phenomena was necessary. From these analyses, (1) accident progressions were identified for inclusion into the APET; (2) branch point probabilities and any attendant parameters were quantified; and (3) the radionuclide releases to the environment from accidents were determined. The phenomena of interest for accident progressions included explosions, fires, a molten glass spill, and the response of the facility confinement system during such challenges. A variety of methodologies, from hand calculations to large system-model codes, were used in the evaluation of these phenomena

Methodological development for selection of significant predictors explaining fatal road accidents.

Science.gov (United States)

Dadashova, Bahar; Arenas-Ramírez, Blanca; Mira-McWilliams, José; Aparicio-Izquierdo, Francisco

2016-05-01

Identification of the most relevant factors for explaining road accident occurrence is an important issue in road safety research, particularly for future decision-making processes in transport policy. However model selection for this particular purpose is still an ongoing research. In this paper we propose a methodological development for model selection which addresses both explanatory variable and adequate model selection issues. A variable selection procedure, TIM (two-input model) method is carried out by combining neural network design and statistical approaches. The error structure of the fitted model is assumed to follow an autoregressive process. All models are estimated using Markov Chain Monte Carlo method where the model parameters are assigned non-informative prior distributions. The final model is built using the results of the variable selection. For the application of the proposed methodology the number of fatal accidents in Spain during 2000-2011 was used. This indicator has experienced the maximum reduction internationally during the indicated years thus making it an interesting time series from a road safety policy perspective. Hence the identification of the variables that have affected this reduction is of particular interest for future decision making. The results of the variable selection process show that the selected variables are main subjects of road safety policy measures. Published by Elsevier Ltd.

A methodology for the transfer of probabilities between accident severity categories

International Nuclear Information System (INIS)

Whitlow, J.D.; Neuhauser, K.S.

1992-01-01

Evaluation of the radiological risks of accidents involving vehicles transporting radioactive materials requires consideration of both accident probability and consequences. The probability that an accident will occur may be estimated from historical accident data for the given mode of transport. In addition to an overall accident rate, information regarding accident severity and the resulting package environments across the range of all credible accidents is needed to determine the potential for a release of radioactive material from the package or for an increase in direct radiation from the package caused by damage to packaging shielding. This information is usually obtained from a variety of sources such as historical data, experimental data, analyses of accident and package environments, and expert opinion. The consequences of an accident depend on a number of factors including the type, quantity, and physical form of radioactive material being transported; the response of the package to accident environments; the fraction of material released from the package; and the dispersion of any released material. One approach for the classification and treatment of transportation accidents in risk analysis divides the complete range of critical accident environments resulting from all credible accidents into some number of accident-severity categories. The types of accident environments that a package may be subjected to in transportation are often classified into the following five groups: impact, fire, crush, puncture, and immersion. A open-quotes criticalclose quotes accident environment is one of a type that could present a plausible threat to a package. Each severity category represents a portion of all credible accidents, and the total of all severity categories covers the complete range of critical accident environments. This approach is used in the risk assessment codes RADTRAN (Neuhauser and Kanipe 1992) and INTERTRAN (Ericsson and Elert 1983)

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.

A calculational methodology for comparing the accident, occupational, and waste-disposal hazards of fusion reactor designs

International Nuclear Information System (INIS)

Fetter, S.

1985-01-01

A methodology has been developed for calculating indices of three classes of radiological hazards: reactor accidents, occupational exposures, and waste-disposal hazards. Radionuclide inventories, biological hazard potentials (BHP), and various dose-related indices are calculated. In the case of reactor accidents, the critical, 50-year and chronic dose are computed, as well as the number of early deaths and illnesses and late cancer fatalities. For occupational exposure, the contact dose rate is calculated for several times after reactor shutdown. In the case of waste-disposal hazards, the intruder dose and the intruder hazard potential (IHP) are calculated. Sample calculations for the MARS reactor design show the usefulness of the methodology in exploring design improvements

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Comorbidity and radiation: methodological aspects of health assessment of persons exposed to the Chornobyl accident factors.

Science.gov (United States)

Nosach, O V

2013-01-01

Comorbidity is one of the most challenging problems of a modern medicine. In a population exposed to the factors of the Chornobyl accident there is an obvious increase in the number of diseases occurring simultaneously against the background of rising prevalence of different classes of chronic medical nosology. The scientific data analysis are presented on the methodological approaches that can be used to create a specialized system for integrated assessment of the health of patients with comorbid disorders. Developing such a system it should be taken into account the trends of changes in the incidence, prevalence and structure of chronic disease, factors and regularities of comorbid disease in the cohorts of Chornobyl accident clean-up workers, evacuees and dwellers of contaminated territories. The system should provide a non-random selection of combinations (clusters) of the most common diseases with serious consequences for the survivors. Nosach O. V., 2013.

Study to improve recriticality evaluation methodology after severe accident (Joint research)

International Nuclear Information System (INIS)

Kugo, Teruhiko; Ishikawa, Makoto; Nagaya, Yasunobu; Yokoyama, Kenji; Fukaya, Yuji; Maruyama, Hiromi; Kondo, Takao; Minato, Hirokazu; Tsuchiya, Akiyuki; Ishii, Yoshihiko; Fujimura, Koji

2014-03-01

The present report summarizes the results of a 2-year cooperative study between JAEA and Hitachi-GE Nuclear Energy in order to contribute to the settlement of the Fukushima-Daiich nuclear power plants which suffered from the severe accident on March 2011. In the Fukushima-Daiich plants, it is considered that the nuclear fuel in the core was melted down by the loss of ultimate heat sink by the station black out after the 2011 off the Pacific coast of Tohoku Earthquake. The position and/or the mass of the melted fuel have not been known yet. Therefore, the possibility of recriticality events and the severity of the postulated recriticality are uncertain. In the present study, the possible scenarios to reach the recriticality events in Fukushima-Daiich were investigated first. Then, the analytical methodology to evaluate the time-dependent recriticality events has been developed by modelling the reactivity insertion rate and the possible feedback according to the recriticality scenarios identified in the first step. The methodology developed here has been equipped as a transient simulation tool, PORCAS, which is operated on a multi-purpose platform for reactor analysis, MARBLE. Finally, the radiation exposure rates by the postulated recriticality events in Fukushima-Daiich were approximately evaluated to estimate the impact to the public environment. (author)

Nuclear fuel cycle facility accident analysis handbook

International Nuclear Information System (INIS)

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

1988-05-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

Consequence analysis of core damage states following severe accidents for the CANDU reactor design

International Nuclear Information System (INIS)

Wahba, N.N.; Kim, Y.T.; Lie, S.G.

1997-01-01

The analytical methodology used to evaluate severe accident sequences is described. The relevant thermal-mechanical phenomena and the mathematical approach used in calculating the timing of the accident progression and source term estimate are summarized. The postulated sever accidents analyzed, in general, mainly differ in the timing to reach and progress through each defined c ore damage state . This paper presents the methodology and results of the timing and steam discharge calculations as well as source term estimate out of containment for accident sequences classified as potentially leading to core disassembly following a small break loss-of-coolant accident (LOCA) scenario as a specific example. (author)

Accident patterns for construction-related workers: a cluster analysis

Science.gov (United States)

Liao, Chia-Wen; Tyan, Yaw-Yauan

2012-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Lee, Su Won

2011-02-15

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

Reactivity accident analysis in MTR cores

International Nuclear Information System (INIS)

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

1987-01-01

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

Uncertainty analysis in calculations of a road accident consequences

International Nuclear Information System (INIS)

Bonnefous, S.; Brenot, J.; Hubert, P.

1995-01-01

This paper develops a concrete situation witch is the search for an evacuation distance in case of a road accident implying a chlorine tank. The methodological aspect is how implementing uncertainty analysis in deterministic models with random parameters. The study demonstrates a great dispersion in the results. It allows to establish satisfactory decision rules and a hierarchy on parameters witch is useful to define priorities in the search for information and to improve the treatment of these parameters. (authors). 8 refs., 1 fig., 2 tabs

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

International Nuclear Information System (INIS)

Nomura, Yasushi

1985-12-01

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

Evaluation of nuclear accidents consequences. Risk assessment methodologies, current status and applications

International Nuclear Information System (INIS)

Rodriguez, J.M.

1996-01-01

General description of the structure and process of the probabilistic methods of assessment the external consequences in the event of nuclear accidents is presented. attention is paid in the interface with Probabilistic Safety Analysis level 3 results (source term evaluation) Also are described key issues in accident consequence evaluation as: effects evaluated (early and late health effects and economic effects due to countermeasures), presentation of accident consequences results, computer codes. Briefly are presented some relevant areas for the applications of Accident Consequence Evaluation

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

International Nuclear Information System (INIS)

Silva, Kampanart; Ishiwatari, Yuki; Takahara, Shogo

2014-01-01

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

Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

Energy Technology Data Exchange (ETDEWEB)

Lee, Han Sul; Heo, Gyun Young [Kyung Hee University, Yongin (Korea, Republic of); Kim, Tae Wan [Incheon National University, Incheon (Korea, Republic of)

2017-03-15

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective.

Application of dynamic probabilistic safety assessment approach for accident sequence precursor analysis: Case study for steam generator tube rupture

International Nuclear Information System (INIS)

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

2017-01-01

The purpose of this research is to introduce the technical standard of accident sequence precursor (ASP) analysis, and to propose a case study using the dynamic-probabilistic safety assessment (D-PSA) approach. The D-PSA approach can aid in the determination of high-risk/low-frequency accident scenarios from all potential scenarios. It can also be used to investigate the dynamic interaction between the physical state and the actions of the operator in an accident situation for risk quantification. This approach lends significant potential for safety analysis. Furthermore, the D-PSA approach provides a more realistic risk assessment by minimizing assumptions used in the conventional PSA model so-called the static-PSA model, which are relatively static in comparison. We performed risk quantification of a steam generator tube rupture (SGTR) accident using the dynamic event tree (DET) methodology, which is the most widely used methodology in D-PSA. The risk quantification results of D-PSA and S-PSA are compared and evaluated. Suggestions and recommendations for using D-PSA are described in order to provide a technical perspective

Methodology of dose calculation for the SRS SAR

International Nuclear Information System (INIS)

Price, J.B.

1991-07-01

The Savannah River Site (SRS) Safety Analysis Report (SAR) covering K reactor operation assesses a spectrum of design basis accidents. The assessment includes estimation of the dose consequences from the analyzed accidents. This report discusses the methodology used to perform the dose analysis reported in the SAR and also includes the quantified doses. Doses resulting from postulated design basis reactor accidents in Chapter 15 of the SAR are discussed, as well as an accident in which three percent of the fuel melts. Doses are reported for both atmospheric and aqueous releases. The methodology used to calculate doses from these accidents as reported in the SAR is consistent with NRC guidelines and industry standards. The doses from the design basis accidents for the SRS reactors are below the limits set for commercial reactors by the NRC and also meet industry criteria. A summary of doses for various postulated accidents is provided

Accidental safety analysis methodology development in decommission of the nuclear facility

Energy Technology Data Exchange (ETDEWEB)

Park, G. H.; Hwang, J. H.; Jae, M. S.; Seong, J. H.; Shin, S. H.; Cheong, S. J.; Pae, J. H.; Ang, G. R.; Lee, J. U. [Seoul National Univ., Seoul (Korea, Republic of)

2002-03-15

Decontamination and Decommissioning (D and D) of a nuclear reactor cost about 20% of construction expense and production of nuclear wastes during decommissioning makes environmental issues. Decommissioning of a nuclear reactor in Korea is in a just beginning stage, lacking clear standards and regulations for decommissioning. This work accident safety analysis in decommissioning of the nuclear facility can be a solid ground for the standards and regulations. For source term analysis for Kori-1 reactor vessel, MCNP/ORIGEN calculation methodology was applied. The activity of each important nuclide in the vessel was estimated at a time after 2008, the year Kori-1 plant is supposed to be decommissioned. And a methodology for risk analysis assessment in decommissioning was developed.

Prediction of road accidents: A Bayesian hierarchical approach.

Science.gov (United States)

Deublein, Markus; Schubert, Matthias; Adey, Bryan T; Köhler, Jochen; Faber, Michael H

2013-03-01

In this paper a novel methodology for the prediction of the occurrence of road accidents is presented. The methodology utilizes a combination of three statistical methods: (1) gamma-updating of the occurrence rates of injury accidents and injured road users, (2) hierarchical multivariate Poisson-lognormal regression analysis taking into account correlations amongst multiple dependent model response variables and effects of discrete accident count data e.g. over-dispersion, and (3) Bayesian inference algorithms, which are applied by means of data mining techniques supported by Bayesian Probabilistic Networks in order to represent non-linearity between risk indicating and model response variables, as well as different types of uncertainties which might be present in the development of the specific models. Prior Bayesian Probabilistic Networks are first established by means of multivariate regression analysis of the observed frequencies of the model response variables, e.g. the occurrence of an accident, and observed values of the risk indicating variables, e.g. degree of road curvature. Subsequently, parameter learning is done using updating algorithms, to determine the posterior predictive probability distributions of the model response variables, conditional on the values of the risk indicating variables. The methodology is illustrated through a case study using data of the Austrian rural motorway network. In the case study, on randomly selected road segments the methodology is used to produce a model to predict the expected number of accidents in which an injury has occurred and the expected number of light, severe and fatally injured road users. Additionally, the methodology is used for geo-referenced identification of road sections with increased occurrence probabilities of injury accident events on a road link between two Austrian cities. It is shown that the proposed methodology can be used to develop models to estimate the occurrence of road accidents for any

A study on the operator's errors of commission (EOC) in accident scenarios of nuclear power plants: methodology development and application

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Whan; Jung, Won Dea; Park, Jin Kyun; Kang, Da Il

2003-04-01

As the concern on the operator's inappropriate interventions, the so-called Errors Of Commission (EOCs), that can exacerbate the plant safety has been raised, much of interest in the identification and analysis of EOC events from the risk assessment perspective has been increased. Also, one of the items in need of improvement for the conventional PSA and HRA that consider only the system-demanding human actions is the inclusion of the operator's EOC events into the PSA model. In this study, we propose a methodology for identifying and analysing human errors of commission that might be occurring from the failures in situation assessment and decision making during accident progressions given an initiating event. In order to achieve this goal, the following research items have been performed: Firstly, we analysed the error causes or situations contributed to the occurrence of EOCs in several incidents/accidents of nuclear power plants. Secondly, limitations of the advanced HRAs in treating EOCs were reviewed, and a requirement for a new methodology for analysing EOCs was established. Thirdly, based on these accomplishments a methodology for identifying and analysing EOC events inducible from the failures in situation assessment and decision making was proposed and applied to all the accident sequences of YGN 3 and 4 NPP which resulted in the identification of about 10 EOC situations.

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

Energy Technology Data Exchange (ETDEWEB)

Jasiulevicius, Audrius

2003-07-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

International Nuclear Information System (INIS)

Oh, Jae Yong; Yun, Jong Il; Kim, Do Sam; Han Chul

2011-01-01

Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

Development of Assessment Methodology of Chemical Behavior of Volatile Iodide under Severe Accident Conditions Using EPICUR Experiments

Energy Technology Data Exchange (ETDEWEB)

Oh, Jae Yong; Yun, Jong Il [KAIST, Daejeon (Korea, Republic of); Kim, Do Sam; Han Chul [Korea Institue of Nuclear Safety, Daejeon (Korea, Republic of)

2011-05-15

Iodine is one of the most important fission products produced in nuclear power plants. Under severe accident condition, iodine exists as a variety of species in the containment such as aqueous iodide, gaseous iodide, iodide aerosol, etc. Following release of iodine from the reactor, mostly in the form of CsI aerosol, volatile iodine can be generated from the containment sump and release to the environment. Especially, volatile organic iodide can be produced from interaction between nonvolatile iodine and organic substances present in the containment. Volatile iodide could significantly influence the alienated residents surrounding the nuclear power plant. In particular, thyroid is vulnerable to radioiodine due to its high accumulation. Therefore, it is necessary for the Korea Institute of Nuclear Safety (KINS) to develop an evaluation model which can simulate iodine behavior in the containment following a severe accident. KINS also needs to make up its methodology for radiological consequence analysis, based on MELCOR-MACCS2 calculation, by coupling a simple iodine model which can conveniently deal with organic iodides. In the long term, such a model can contribute to develop an accident source term, which is one of urgent domestic needs. Our strategy for developing the model is as follows: 1. Review the existing methodologies, 2. Develop a simple stand-alone model, 3. Validate the model using ISTP-EPICUR (Experimental Program on Iodine Chemistry under Radiation) and OECD-BIP (Behavior of Iodine Project) experimental data. In this paper we present the context of development and validation of our model named RAIM (Radio-active iodine chemistry model)

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

International Nuclear Information System (INIS)

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

2007-01-01

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

Risk assessment of maintenance operations: the analysis of performing task and accident mechanism.

Science.gov (United States)

Carrillo-Castrillo, Jesús A; Rubio-Romero, Juan Carlos; Guadix, Jose; Onieva, Luis

2015-01-01

Maintenance operations cover a great number of occupations. Most small and medium-sized enterprises lack the appropriate information to conduct risk assessments of maintenance operations. The objective of this research is to provide a method based on the concepts of task and accident mechanisms for an initial risk assessment by taking into consideration the prevalence and severity of the maintenance accidents reported. Data were gathered from 11,190 reported accidents in maintenance operations in the manufacturing sector of Andalusia from 2003 to 2012. By using a semi-quantitative methodology, likelihood and severity were evaluated based on the actual distribution of accident mechanisms in each of the tasks. Accident mechanisms and tasks were identified by using those variables included in the European Statistics of Accidents at Work methodology. As main results, the estimated risk of the most frequent accident mechanisms identified for each of the analysed tasks is low and the only accident mechanisms with medium risk are accidents when lifting or pushing with physical stress on the musculoskeletal system in tasks involving carrying, and impacts against objects after slipping or stumbling for tasks involving movements. The prioritisation of public preventive actions for the accident mechanisms with a higher estimated risk is highly recommended.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Methodology for risk analysis of nuclear installations

International Nuclear Information System (INIS)

Vasconcelos, Vanderley de; Senne Junior, Murillo; Jordao, Elizabete

2002-01-01

Both the licensing standards for general uses in nuclear facilities and the specific ones require a risk assessment during their licensing processes. The risk assessment is carried out through the estimation of both probability of the occurrence of the accident, and their magnitudes. This is a complex task because the great deal of potential hazardous events that can occur in nuclear facilities difficult the statement of the accident scenarios. There are also many available techniques to identify the potential accidents, estimate their probabilities, and evaluate their magnitudes. In this paper is presented a new methodology that systematizes the risk assessment process, and orders the accomplishment of their several steps. (author)

The Analysis of Uncontrolled CEA Bank withdrawal from low power Accident for APR1400 Using KNAP

International Nuclear Information System (INIS)

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

2006-01-01

In Korea, the nuclear industries such as fuel manufacturer, the architect engineer and the utility, have been using the methodologies and codes of vendors, such as Westinghouse(WH), Combustion Engineering(CE), for the safety analyses of nuclear power plants. Consequently the industries have kept up the many organizations to operate the methodologies and to maintain the codes for each vendor. It may cause difficulty to improve the safety analyses efficiency and technology related. So, the necessity another of methodologies and code systems applicable to Non- LOCA, beyond design basis accident and performance analyses for all types of pressurized water reactor (PWR) has been raised. As the first requirement, the best-estimate codes were required for applicable wider application area and realistic behavior prediction of power plants with various and sophisticated functions. After the review on several candidates, RETRAN-3D has been chosen as a system analysis code. The draft version of the methodology was developed based on the references for the general purpose, and modified to apply it to specific plants in Korea. As a part of the feasibility estimation for the methodology and code system, uncontrolled CEA bank withdrawal from low power accident for the Advanced Power Reactor 1400(APR1400) was selected to verify the feasibility using the RETRAN-3D. And the results were compared with the Standard Safety Analysis Reports (SSAR) of APR1400

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

DEVELOPMENT OF METHODOLOGY FOR TRAFFIC ACCIDENT FORECASTING AT VARIOUS TYPICAL URBAN AREAS

OpenAIRE

D. V. Kapsky

2012-01-01

The paper provides investigation results pertaining to development of methodology for forecasting traffic accidents using a “conflict zone” method that considers potential danger for two typical urban areas, namely: signaled crossings and bumps that are made in the areas of zebra crossings and it also considers various types and kinds of conflicts. The investigations have made it possible to obtain various indices of threshold sensitivity in respect of  potential risks  and in relation to tra...

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

International Nuclear Information System (INIS)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang

2005-01-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

Analysis of the main steam line break accident with loss of offsite power using the fully coupled RELAP5/PANTHER/COBRA code package

Energy Technology Data Exchange (ETDEWEB)

Ruben Van Parys; Sandrine Bosso; Christophe Schneidesch; Jinzhao Zhang [Nuclear Department, Suez-Tractebel Engineering, avenue Ariane 5, B-1200 Brussels (Belgium)

2005-07-01

Full text of publication follows: A coupled thermal hydraulics-neutronics code package (RELAP5/PANTHER/COBRA) has been qualified for accident analysis at Tractebel Engineering. In the TE coupled code package, the best estimate thermal-hydraulic system code, RELAP5/MOD2.5, is coupled with the full three-dimensional reactor core kinetics code, PANTHER, via a dynamic data exchange control and processing tool, TALINK. An interface between PANTHER code and the sub-channel thermal-hydraulic analysis code COBRA-IIIC is developed in order to perform online calculation of Departure from Nucleate Boiling Ratio (DNBR). The TE coupled code package has been applied to develop a MSLB accident analysis methodology using the TE deterministic bounding approach. The methodology has been applied for MSLB accident analysis in support of licensing of the power up-rate and steam generator replacement of the Doel 2 plant. The results of coupled thermal-hydraulic and neutronic analysis of SLB show that there exists an important margin in the traditional FSAR MSLB accident analysis. As a specific licensing requirement, the main steam line break accident with loss of offsite power has to be analyzed. In the standard methodology with the coupled RELAP5/PANTHER code, and some corrective methods has to be taken in order to overcome the limitations due to the close-channel T/H model in PANTHER at low flow conditions. The results show that the steam line break accident with loss of offsite power is far less limiting. In order to verify the effect of the cross-flow at low flow conditions, the fully dynamic coupling of RELAP5/PANTHER/COBRA code package is used for reanalysis of this case, in which the PANTHER close-channel T/H model is replaced by the COBRA sub-channel T/H model with crossflow option. It has been demonstrated that, although the consideration of cross-flow in this challenging situation may lead to higher core return to power and slightly lower DNBR than in the standard methodology

PERSPECTIVES ON A DOE CONSEQUENCE INPUTS FOR ACCIDENT ANALYSIS APPLICATIONS

International Nuclear Information System (INIS)

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

2008-01-01

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

Accident Analysis and Highway Safety

Directory of Open Access Journals (Sweden)

Omar Noorliyana

2017-01-01

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

Analysis methodology for RBMK-1500 core safety and investigations on corium coolability during a LWR severe accident

International Nuclear Information System (INIS)

Jasiulevicius, Audrius

2003-01-01

This thesis presents the work involving two broad aspects within the field of nuclear reactor analysis and safety. These are: - development of a fully independent reactor dynamics and safety analysis methodology of the RBMK-1500 core transient accidents and - experiments on the enhancement of coolability of a particulate bed or a melt pool due to heat removal through the control rod guide tubes. The first part of the thesis focuses on the development of the RBMK-1500 analysis methodology based on the CORETRAN code package. The second part investigates the issue of coolability during severe accidents in LWR type reactors: the coolability of debris bed and melt pool for in-vessel and ex-vessel conditions. The first chapter briefly presents the status of developments in both the RBMK-1500 core analysis and the corium coolability areas. The second chapter describes the generation of the RBMK-1500 neutron cross section data library with the HELIOS code. The cross section library was developed for the whole range of the reactor conditions. The results of the benchmarking with the WIMS-D4 code and validation against the RBMK Critical Facility experiments is also presented here. The HELIOS generated neutron cross section data library provides a close agreement with the WIMS-D4 code results. The validation against the data from the Critical Experiments shows that the HELIOS generated neutron cross section library provides excellent predictions for the criticality, axial and radial power distribution, control rod reactivity worths and coolant reactivity effects, etc. The reactivity effects of voiding for the system, fuel assembly and additional absorber channel are underpredicted in the calculations using the HELIOS code generated neutron cross sections. The underprediction, however, is much less than that obtained when the WIMS-D4 code generated cross sections are employed. The third chapter describes the work, performed towards the accurate prediction, assessment and

PSA methodology development and application in Japan

International Nuclear Information System (INIS)

Kazuo Sato; Toshiaki Tobioka; Kiyoharu Abe

1987-01-01

The outlines of Japanese activities on development and application of probabilistic safety assessment (PSA) methodologies are described. First the activities on methodology development are described for system reliability analysis, operational data analysis, core melt accident analysis, environmental consequence analysis and seismic risk analysis. Then the methodoligy application examples by the regulatory side and the industry side are described. (author)

Accident analysis for PRC-II reactor

International Nuclear Information System (INIS)

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

1997-12-01

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

Uncertainty and sensitivity analysis in nuclear accident consequence assessment

International Nuclear Information System (INIS)

Karlberg, Olof.

1989-01-01

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

Analysis of Fukushima Daiichi Accident Using HFACS

International Nuclear Information System (INIS)

Mohamed, Saeed Almheiri

2013-01-01

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

Analysis of Fukushima Daiichi Accident Using HFACS

Energy Technology Data Exchange (ETDEWEB)

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

2013-10-15

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

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

Science.gov (United States)

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

2016-01-01

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

Nuclear accidents and epidemiology

International Nuclear Information System (INIS)

1987-01-01

A consultation on epidemiology related to the Chernobyl accident was held in Copenhagen in May 1987 as a basis for concerted action. This was followed by a joint IAEA/WHO workshop in Vienna, which reviewed appropriate methodologies for possible long-term effects of radiation following nuclear accidents. The reports of these two meetings are included in this volume, and cover the subjects: 1) Epidemiology related to the Chernobyl nuclear accident. 2) Appropriate methodologies for studying possible long-term effects of radiation on individuals exposed in a nuclear accident. Figs and tabs

Progress in the development of methodology for fusion safety systems studies

International Nuclear Information System (INIS)

Ho, S.K.; Cambi, G.; Ciattaglia, S.; Fujii-e, Y.; Seki, Y.

1994-01-01

The development of fusion safety systems-study methodology, including the aspects of schematic classification of overall fusion safety system, qualitative assessment of fusion system for identification of critical accident scenarios, quantitative analysis of accident consequences and risk for safety design evaluation, and system-level analysis of accident consequences and risk for design optimization, by a consortium of international efforts is presented. The potential application of this methodology into reactor design studies will facilitate the systematic assessment of safety performance of reactor designs and enhance the impacts of safety considerations on the selection of design configurations

Accident sequence analysis for a BWR [Boiling Water Reactor] during low power and shutdown operations

International Nuclear Information System (INIS)

Whitehead, D.W.; Hake, T.M.

1990-01-01

Most previous Probabilistic Risk Assessments have excluded consideration of accidents initiated in low power and shutdown modes of operation. A study of the risk associated with operation in low power and shutdown is being performed at Sandia National Laboratories for a US Boiling Water Reactor (BWR). This paper describes the proposed methodology for the analysis of the risk associated with the operation of a BWR during low power and shutdown modes and presents preliminary information resulting from the application of the methodology. 2 refs., 2 tabs

Analysis of severe accidents in pressurized heavy water reactors

International Nuclear Information System (INIS)

2008-06-01

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

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.

Neutronics and thermal-hydraulics coupling: some contributions toward an improved methodology to simulate the initiating phase of a severe accident in a sodium fast reactor

International Nuclear Information System (INIS)

Guyot, Maxime

2014-01-01

This project is dedicated to the analysis and the quantification of bias corresponding to the computational methodology for simulating the initiating phase of severe accidents on Sodium Fast Reactors. A deterministic approach is carried out to assess the consequences of a severe accident by adopting best estimate design evaluations. An objective of this deterministic approach is to provide guidance to mitigate severe accident developments and re-criticalities through the implementation of adequate design measures. These studies are generally based on modern simulation techniques to test and verify a given design. The new approach developed in this project aims to improve the safety assessment of Sodium Fast Reactors by decreasing the bias related to the deterministic analysis of severe accident scenarios. During the initiating phase, the subassembly wrapper tubes keep their mechanical integrity. Material disruption and dispersal is primarily one-dimensional. For this reason, evaluation methodology for the initiating phase relies on a multiple-channel approach. Typically a channel represents an average pin in a subassembly or a group of similar subassemblies. In the multiple-channel approach, the core thermal-hydraulics model is composed of 1 or 2 D channels. The thermal-hydraulics model is coupled to a neutronics module to provide an estimate of the reactor power level. In this project, a new computational model has been developed to extend the initiating phase modeling. This new model is based on a multi-physics coupling. This model has been applied to obtain information unavailable up to now in regards to neutronics and thermal-hydraulics models and their coupling. (author) [fr

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

International Nuclear Information System (INIS)

Marrel, A.; Marie, N.; De Lozzo, M.

2015-01-01

Within the framework of the generation IV Sodium Fast Reactors, the safety in case of severe accidents is assessed. From this statement, CEA has developed a new physical tool to model the accident initiated by the Total Instantaneous Blockage (TIB) of a sub-assembly. This TIB simulator depends on many uncertain input parameters. This paper aims at proposing a global methodology combining several advanced statistical techniques in order to perform a global sensitivity analysis of this TIB simulator. The objective is to identify the most influential uncertain inputs for the various TIB outputs involved in the safety analysis. The proposed statistical methodology combining several advanced statistical techniques enables to take into account the constraints on the TIB simulator outputs (positivity constraints) and to deal simultaneously with various outputs. To do this, a space-filling design is used and the corresponding TIB model simulations are performed. Based on this learning sample, an efficient constrained Gaussian process metamodel is fitted on each TIB model outputs. Then, using the metamodels, classical sensitivity analyses are made for each TIB output. Multivariate global sensitivity analyses based on aggregated indices are also performed, providing additional valuable information. Main conclusions on the influence of each uncertain input are derived. - Highlights: • Physical-statistical tool for Sodium Fast Reactors TIB accident. • 27 uncertain parameters (core state, lack of physical knowledge) are highlighted. • Constrained Gaussian process efficiently predicts TIB outputs (safety criteria). • Multivariate sensitivity analyses reveal that three inputs are mainly influential. • The type of corium propagation (thermal or hydrodynamic) is the most influential

Identification of reference accident scenarios in SEVESO establishments

International Nuclear Information System (INIS)

Delvosalle, C.; Fievez, C.; Pipart, A.; Fabrega, J. Casal; Planas, E.; Christou, M.; Mushtaq, F.

2005-01-01

In the frame of the ESREL special session on ARAMIS project, this paper aims at presenting the work carried out in the first Work Package, 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. A first part of the work aims at building 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. This work is coupled with an historical analysis of accidents. In a second part, influence of safety devices and policies will be considered, in order to build a Methodology for the Identification of Reference Accident Scenarios (MIRAS). This last one will take into account safety systems and lead to obtain more realistic scenarios

CONTAINMENT ANALYSIS METHODOLOGY FOR TRANSPORT OF BREACHED CLAD ALUMINUM SPENT FUEL

Energy Technology Data Exchange (ETDEWEB)

Vinson, D.

2010-07-11

Aluminum-clad, aluminum-based spent nuclear fuel (Al-SNF) from foreign and domestic research reactors (FRR/DRR) is being shipped to the Savannah River Site and placed in interim storage in a water basin. To enter the United States, a cask with loaded fuel must be certified to comply with the requirements in the Title 10 of the U.S. Code of Federal Regulations, Part 71. The requirements include demonstration of containment of the cask with its contents under normal and accident conditions. Many Al-SNF assemblies have suffered corrosion degradation in storage in poor quality water, and many of the fuel assemblies are 'failed' or have through-clad damage. A methodology was developed to evaluate containment of Al-SNF even with severe cladding breaches for transport in standard casks. The containment analysis methodology for Al-SNF is in accordance with the methodology provided in ANSI N14.5 and adopted by the U. S. Nuclear Regulatory Commission in NUREG/CR-6487 to meet the requirements of 10CFR71. The technical bases for the inputs and assumptions are specific to the attributes and characteristics of Al-SNF received from basin and dry storage systems and its subsequent performance under normal and postulated accident shipping conditions. The results of the calculations for a specific case of a cask loaded with breached fuel show that the fuel can be transported in standard shipping casks and maintained within the allowable release rates under normal and accident conditions. A sensitivity analysis has been conducted to evaluate the effects of modifying assumptions and to assess options for fuel at conditions that are not bounded by the present analysis. These options would include one or more of the following: reduce the fuel loading; increase fuel cooling time; reduce the degree of conservatism in the bounding assumptions; or measure the actual leak rate of the cask system. That is, containment analysis for alternative inputs at fuel-specific conditions and

HANARO thermal hydraulic accident analysis

Energy Technology Data Exchange (ETDEWEB)

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

1996-06-01

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

Prototype application of best estimate and uncertainty safety analysis methodology to large LOCA analysis

International Nuclear Information System (INIS)

Luxat, J.C.; Huget, R.G.

2001-01-01

Development of a methodology to perform best estimate and uncertainty nuclear safety analysis has been underway at Ontario Power Generation for the past two and one half years. A key driver for the methodology development, and one of the major challenges faced, is the need to re-establish demonstrated safety margins that have progressively been undermined through excessive and compounding conservatism in deterministic analyses. The major focus of the prototyping applications was to quantify the safety margins that exist at the probable range of high power operating conditions, rather than the highly improbable operating states associated with Limit of the Envelope (LOE) assumptions. In LOE, all parameters of significance to the consequences of a postulated accident are assumed to simultaneously deviate to their limiting values. Another equally important objective of the prototyping was to demonstrate the feasibility of conducting safety analysis as an incremental analysis activity, as opposed to a major re-analysis activity. The prototype analysis solely employed prior analyses of Bruce B large break LOCA events - no new computer simulations were undertaken. This is a significant and novel feature of the prototyping work. This methodology framework has been applied to a postulated large break LOCA in a Bruce generating unit on a prototype basis. This paper presents results of the application. (author)

Status of science and technology with respect of preparation and evaluation of accident analyses and the use of analysis simulators

International Nuclear Information System (INIS)

Pointner, Winfried; Cuesta Morales, Alejandra; Draeger, Peer; Hartung, Juergen; Jakubowski, Zygmunt; Meyer, Gerhard; Palazzo, Simone; Moner, Guim Pallas; Perin, Yann; Pasichnyk, Ihor

2014-07-01

The scope of the work was to elaborate the prerequisites for short term accident analyses including recommendations for the application of new methodologies and computational procedures and technical aspects of safety evaluation. The following work packages were performed: Knowledge base for best estimate accident analyses; analytical studies on the PWR plant behavior in case of multiple safety system failures; extension and maintenance of the data base for plant specific analysis simulators.

Application of forensic image analysis in accident investigations.

Science.gov (United States)

Verolme, Ellen; Mieremet, Arjan

2017-09-01

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

Canister storage building design basis accident analysis documentation

International Nuclear Information System (INIS)

KOPELIC, S.D.

1999-01-01

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

The Need to introduce CFD Methodology in Analyze Hydrogen Distribution for Postulated Severe Accidents

Energy Technology Data Exchange (ETDEWEB)

Na, Hanbee; Park, Sukyung; Kim, Kyuntae [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of); Lee, Jongkwang [Hanbat National University, Daejeon (Korea, Republic of); Kwon, Sejin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2015-10-15

The regulatory requirements for combustible gas control systems in Korea is that mean hydrogen mole fraction shall be lower than 10 %, containment integrity shall be kept from combustion of hydrogen, and detonation and global fast turbulent combustion shall be avoided. KHNP provided some analysis which show hydrogen mole fraction is less than 10 % and detonation and global fast turbulence combustion are avoided for postulated severe accident events which covered over 90 % of CDF (core damage frequency) for each NPP. The results were from MAAP code that can simulate from the initiation of the accidents to hydrogen distribution inside containments. It is a Lumped-Parameter codes in which the transport of energy and mass is possible in only predetermined one direction. Therefore, there has been a long-history dispute whether one-dimensional LP codes could simulate the transportation of hydrogen accurately. For example, KHNP made a MAAP model to simulate hydrogen distribution in KSNP (Korean Standard Nuclear Plants), and the containment free volume is divided into 27 nodes in which it is assumed all the properties like each molecule mole fraction and temperate are uniform in each node. In addition, the maximum volume size of them is over 22,000 m{sup 3}, and it is not quite confident that the mole fraction of each molecules and temperature are uniform in the big size space. As for the stress test results of the Wolsong 1, civil experts asked KHNP to conduct hydrogen distribution analysis using Computational Fluid Dynamics (CFD) methodology, and if needed to install hydrogen ignitors in Wolsong 1 NPP. As a reviewer for KHNP's post actions to the Stress Test, the author also asked KHNP to do CFD analysis of hydrogen distribution, and KHNP finally agreed to analyze it using CFD by 2017. KHNP submitted a Shin-hanul 1 and 2 Operation License application in 2015, and the author also asked it to do CFD analysis to simulate hydrogen distribution for Shin-hanul 1 and 2

Development of a DNBR evaluation method for the CEA ejection accident in SMART core

Energy Technology Data Exchange (ETDEWEB)

Hwang, Dae Hyun; Yoo, Y. J.; In, W. K.; Chang, M. H. [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-12-01

A methodology applicable to the analysis of the CEA ejection accident in SMART is developed for the evaluation of the fraction of fuel failure caused by DNB. The transient behavior of the core thermal-hydraulic conditions is calculated by the subchannel analysis code MATRA. The minimum DNBR during the accident is calculated by KRB-1 CHF correlation considering the 1/8 symmetry of hot assembly. The variation of hot assembly power during the accident is simulated by the LTC(Limiting transient Curve) which is determined from the analysis of power distribution data resulting from the three-dimensional core dynamics calculations. The initial condition of the accident is determined by considering LOC(Limiting Conditions for Operation) of SMART core. Two different methodologies for the evaluation of DNB failure rate are established; a deterministic method based on the DNB envelope, and a probabilistic method based on the DNB probability of each fuel rod. The methodology developed in this study is applied to the analysis of CEA ejection accident in the preliminary design core of SMART. As the result, the fractions of DNB fuel failure by the deterministic method and the probabilistic method are calculated as 38.7% and 7.8%, respectively. 16 refs., 16 figs., 5 tabs. (Author)

Physical data generation methodology for return-to-power steam line break analysis

Energy Technology Data Exchange (ETDEWEB)

Zee, Sung Kyun; Lee, Chung Chan; Lee, Chang Kue [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-02-01

Current methodology to generate physics data for steamline break accident analysis of CE-type nuclear plant such as Yonggwang Unit 3 is valid only if the core reactivity does not reach the criticality after shutdown. Therefore, the methodology requires tremendous amount of net scram worth, specially at the end of the cycle when moderator temperature coefficient is most negative. Therefore, we need a new methodology to obtain reasonably conservation physics data, when the reactor returns to power condition. Current methodology used ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Current methodology uses ROCS which include only closed channel model. But it is well known that the closed channel model estimates the core reactivity too much negative if core flow rate is low. Therefore, a conservative methodology is presented which utilizes open channel 3D HERMITE model. Return-to-power reactivity credit is produced to assist the reactivity table generated by closed channel model. Other data includes hot channel axial power shape, peaking factor and maximum quality for DNBR analysis. It also includes pin census for radiological consequence analysis. 48 figs., 22 tabs., 18 refs. (Author) .new.

Analysis and research status of severe core damage accidents

International Nuclear Information System (INIS)

1984-03-01

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

Preliminary safety analysis of the PWR with accident-tolerant fuels during severe accident conditions

International Nuclear Information System (INIS)

Wu, Xiaoli; Li, Wei; Wang, Yang; Zhang, Yapei; Tian, Wenxi; Su, Guanghui; Qiu, Suizheng; Liu, Tong; Deng, Yongjun; Huang, Heng

2015-01-01

Highlights: • Analysis of severe accident scenarios for a PWR fueled with ATF system is performed. • A large-break LOCA without ECCS is analyzed for the PWR fueled with ATF system. • Extended SBO cases are discussed for the PWR fueled with ATF system. • The accident-tolerance of ATF system for application in PWR is illustrated. - Abstract: Experience gained in decades of nuclear safety research and previous nuclear accidents direct to the investigation of passive safety system design and accident-tolerant fuel (ATF) system which is now becoming a hot research point in the nuclear energy field. The ATF system is aimed at upgrading safety characteristics of the nuclear fuel and cladding in a reactor core where active cooling has been lost, and is preferable or comparable to the current UO 2 –Zr system when the reactor is in normal operation. By virtue of advanced materials with improved properties, the ATF system will obviously slow down the progression of accidents, allowing wider margin of time for the mitigation measures to work. Specifically, the simulation and analysis of a large break loss of coolant accident (LBLOCA) without ECCS and extended station blackout (SBO) severe accident are performed for a pressurized water reactor (PWR) loaded with ATF candidates, to reflect the accident-tolerance of ATF

MELCOR analysis of the TMI-2 accident

International Nuclear Information System (INIS)

Boucheron, E.A.

1990-01-01

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

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.

Canister storage building design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

KOPELIC, S.D.

1999-02-25

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

Development of Database for Accident Analysis in Indian Mines

Science.gov (United States)

Tripathy, Debi Prasad; Guru Raghavendra Reddy, K.

2016-10-01

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

3-D rod ejection analysis using a conservative methodology

Energy Technology Data Exchange (ETDEWEB)

Park, Min Ho; Park, Jin Woo; Park, Guen Tae; Um, Kil Sup; Ryu, Seok Hee; Lee, Jae Il; Choi, Tong Soo [KEPCO, Daejeon (Korea, Republic of)

2016-05-15

The point kinetics model which simplifies the core phenomena and physical specifications is used for the conventional rod ejection accident analysis. The point kinetics model is convenient to assume conservative core parameters but this simplification loses large amount of safety margin. The CHASER system couples the three-dimensional core neutron kinetics code ASTRA, the sub-channel analysis code THALES and the fuel performance analysis code FROST. The validation study for the CHASER system is addressed using the NEACRP three-dimensional PWR core transient benchmark problem. A series of conservative rod ejection analyses for the APR1400 type plant is performed for both hot full power (HFP) and hot zero power (HZP) conditions to determine the most limiting cases. The conservative rod ejection analysis methodology is designed to properly consider important phenomena and physical parameters.

Methodology of radionuclides dis incorporation in people related to nuclear and radiological accidents

International Nuclear Information System (INIS)

Jimenez F, E. A.

2014-01-01

In this paper a classification of the radiological and nuclear accidents is presented, describing which the activities are, where they have occurred, their incidence and the learned lessons in these successes. The radiological accidents in which radioactive materials intervene can occur anywhere, and they are related to no controlled dangerous sources (abandoned, lost, stolen, or found sources), improper use of dangerous industrial and medical sources, exposition and contamination of people in general by an unknown origin, serious over expositions, menaces and willful misconduct, emergencies during transportation of radioactive material. A person can receive a dose of radiation from an external source, because of radioactive material placed on skin or on equipment, or because of ingestion or inhalation of radiological particles. The ingestion or the inhalation of radioactive material can cause an internal dose to the whole body or to a specific organ during a period of time. That is why a description of the processes of incorporation, the stages of incorporation and a description of the biokinetic models are also realized to understand the ingestion, transference and the excretion of the radioactive elements. In order to offer help to a victim of internal contamination, the dosimetric and medical diagnosis is very important. The most important techniques of dosimetric diagnosis are the dosimetry in vivo (cytogenetics and the counting in vivo of the whole body) and the bioassays. These techniques allow obtain data such as the radionuclide, the target organ, the absorbed dose, etc. At the same time, the doctor in charge must be attentive to the patients symptoms and their manifestation time, since they are an indicator, first, the patient suffered an irradiation, and second, of the range esteem of the received radiation dose. These are the parameters that are useful as criterion to decide if a person has to receive some treatment and select the methodologies that

Progress in methodology for probabilistic assessment of accidents: timing of accident sequences

International Nuclear Information System (INIS)

Lanore, J.M.; Villeroux, C.; Bouscatie, F.; Maigret, N.

1981-09-01

There is an important problem for probabilistic studies of accident sequences using the current event tree techniques. Indeed this method does not take into account the dependence in time of the real accident scenarios, involving the random behaviour of the systems (lack or delay in intervention, partial failures, repair, operator actions ...) and the correlated evolution of the physical parameters. A powerful method to perform the probabilistic treatment of these complex sequences (dynamic evolution of systems and associated physics) is Monte-Carlo simulation, very rare events being treated with the help of suitable weighting and biasing techniques. As a practical example the accident sequences related to the loss of the residual heat removal system in a fast breeder reactor has been treated with that method

Safety analysis of accident localization system

International Nuclear Information System (INIS)

1999-01-01

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

Dutch in-depth accident investigation: first experiences and analysis results for motorcycles and mopeds

NARCIS (Netherlands)

Mooi, H.G.; Galliano, F.

2001-01-01

In September 1999 the Dutch Accident Research Team (DART) within TNO Automotive started with the in-depth investigation of traffic accidents. In this paper, the methodology, working procedures and experiences of the team are described and explained in detail. Furthermore, an elaborate description of

An approach to accidents modeling based on compounds road environments.

Science.gov (United States)

Fernandes, Ana; Neves, Jose

2013-04-01

The most common approach to study the influence of certain road features on accidents has been the consideration of uniform road segments characterized by a unique feature. However, when an accident is related to the road infrastructure, its cause is usually not a single characteristic but rather a complex combination of several characteristics. The main objective of this paper is to describe a methodology developed in order to consider the road as a complete environment by using compound road environments, overcoming the limitations inherented in considering only uniform road segments. The methodology consists of: dividing a sample of roads into segments; grouping them into quite homogeneous road environments using cluster analysis; and identifying the influence of skid resistance and texture depth on road accidents in each environment by using generalized linear models. The application of this methodology is demonstrated for eight roads. Based on real data from accidents and road characteristics, three compound road environments were established where the pavement surface properties significantly influence the occurrence of accidents. Results have showed clearly that road environments where braking maneuvers are more common or those with small radii of curvature and high speeds require higher skid resistance and texture depth as an important contribution to the accident prevention. Copyright © 2013 Elsevier Ltd. All rights reserved.

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

International Nuclear Information System (INIS)

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

1994-01-01

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

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.

Accident analysis and DOE criteria

International Nuclear Information System (INIS)

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

1982-01-01

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

Fatal accidents analysis in Peruvian mining industry

International Nuclear Information System (INIS)

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

2010-01-01

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

An analysis of the Three Mile Island accident

International Nuclear Information System (INIS)

Brooks, G.L.; Siddal, E.

1980-09-01

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

Development of a methodology for accident causation research

Science.gov (United States)

1983-06-01

The obj ective of this study was to fully develop and apply a me thodology to : study accident causation, uhich was outlined in a previous study . " Causal" factors : are those pre-crash factors, which are statistically related to the accident rate :...

Analysis of labor accidents in Brazil, 2004-2007

OpenAIRE

Alves, Everton Fernando

2010-01-01

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

Some Examples of Accident Analyses for RB Reactor

International Nuclear Information System (INIS)

Pesic, M.

2002-01-01

The RB reactor is heavy water critical assembly operated in the Vinca Institute of Nuclear Sciences, Belgrade, Yugoslavia, since April 1959. The first Safety Analysis Report of the RB critical assembly was prepared in 1961/62. But, the first accidental analysis was done in late 1958 in aim the examine power transient and total equivalent doses received by the staff during the reactivity accident occurred on October 15, 1958. Since 1960, the RB reactor is modified few times. Beside initial natural uranium metal fuel rods, new fuel (TVR-S types) from 2% enriched metal uranium and 80% enriched UO 2 were available since 1962 and 1976, respectively. Also, modifications in control and safety systems of the reactor were done occasionally. Special reactor cores were created using all three types of fuel elements, among them, the coupled fast-thermal ones. Nuclear Safety Committee of the Vinca Institute, an independent regulatory body approved for usage all these modifications of the RB reactor. For those decisions of the Committee, the Preliminary Safety Analysis Reports were prepared that, beside proposed technical modifications and new regulation rules had included analyses of various possible accidents. Special attention is given and new methodology was proposed for thoroughly analyses of design based accidents related to coupled fast-thermal cores, that include reactor central zones filled by fuel elements without moderator. In these accidents, during assumed flooding of the fast zone by moderator, a very high reactivity could be inserted in the system with very high reactivity rate. It was necessary to provide that the safety system of the reactor had fast response to that accident and had enough high (negative) reactivity to shut down the reactor timely. In this paper, a brief overview of some accidents, methodology and computation tools used for the accident analyses at RB reactor are given. (author)

Development of integrated accident management assessment technology

International Nuclear Information System (INIS)

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

2002-04-01

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

Development of Non-LOCA Safety Analysis Methodology with RETRAN-3D and VIPRE-01/K

International Nuclear Information System (INIS)

Kim, Yo-Han; Cheong, Ae-Ju; Yang, Chang-Keun

2004-01-01

Korea Electric Power Research Institute has launched a project to develop an in-house non-loss-of-coolant-accident analysis methodology to overcome the hardships caused by the narrow analytical scopes of existing methodologies. Prior to the development, some safety analysis codes were reviewed, and RETRAN-3D and VIPRE-01 were chosen as the base codes. The codes have been modified to improve the analytical capabilities required to analyze the nuclear power plants in Korea. The methodologies of the vendors and the Electric Power Research Institute have been reviewed, and some documents of foreign utilities have been used to compensate for the insufficiencies. For the next step, a draft methodology for pressurized water reactors has been developed and modified to apply to Westinghouse-type plants in Korea. To verify the feasibility of the methodology, some events of Yonggwang Units 1 and 2 have been analyzed from the standpoints of reactor coolant system pressure and the departure from nucleate boiling ratio. The results of the analyses show trends similar to those of the Final Safety Analysis Report

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

NARCIS (Netherlands)

Bijleveld, F. D.

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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

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

2000-01-01

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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

CROWE, R.D.

1999-01-01

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

Accident analysis in nuclear power plants

International Nuclear Information System (INIS)

Silva, D.E. da

1981-01-01

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

Gas-cooled reactor safety and accident analysis

International Nuclear Information System (INIS)

1985-12-01

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

Cold Vacuum Drying Facility Design Basis Accident Analysis Documentation

International Nuclear Information System (INIS)

PIEPHO, M.G.

1999-01-01

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

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

International Nuclear Information System (INIS)

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

2016-01-01

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

Accident Sequence Evaluation Program: Human reliability analysis procedure

International Nuclear Information System (INIS)

Swain, A.D.

1987-02-01

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

Energy Analysis of Road Accidents Based on Close-Range Photogrammetry

Directory of Open Access Journals (Sweden)

Alejandro Morales

2015-11-01

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

A three-dimensional pin-wise analysis for CEA ejection accident

Energy Technology Data Exchange (ETDEWEB)

Park, Guen-Tae; Park, Min-Ho; Park, Jin-Woo; Um, Kil-Sup; Choi, Tong-Soo [KEPCO NF, Daejeon (Korea, Republic of)

2016-10-15

The ejection of a control element assembly (CEA) with high reactivity worth causes the sudden insertion of reactivity into the core. Immediately after the CEA ejection, the nuclear power of the reactor dramatically increases in an exponential behavior until the doppler effect becomes important and turns the reactivity balance and power down to lower levels. The 3-D CEA ejection analysis methodology has been developed using the multi-dimensional code coupling system, CHASER, which couples three dimensional core neutron kinetics code ASTRA, subchannel analysis code THALES, and fuel performance analysis code FROST using message passing interface (MPI). This paper presents the pin-by-pin level analysis result with the 3-D CEA ejection analysis methodology using the CHASER. The pin-by-pin level analysis consists of DNBR, enthalpy and Pellet/Clad Mechanical Interaction (PCMI) analysis. All the evaluations are simulated for APR1400 plant loaded with PLUS7 fuel. In this paper, the pin-by-pin analysis using the multidimensional core transient code, CHASER, is presented with respect to enthalpy, DNBR and PCMI for APR1400 plant loaded with PLUS7 fuel. For the pin-by-pin enthalpy and DNBR analysis, the quarter core for HFP case or 15 - 20 assemblies around the most severe assembly for part powers or HZP cases are selected. And PCMI calculation is performed for all the rods in the whole core during a conservative time period. The pin-by-pin analysis results show that the regulatory guidelines of CEA ejection accident are satisfied.

Severe accident analysis and management in nuclear fuel cycle facilities

International Nuclear Information System (INIS)

Golshan, Mina

2013-01-01

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

A cluster analysis on road traffic accidents using genetic algorithms

Science.gov (United States)

Saharan, Sabariah; Baragona, Roberto

2017-04-01

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

Limitations of systemic accident analysis methods

Directory of Open Access Journals (Sweden)

Casandra Venera BALAN

2016-12-01

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

Core melt progression and consequence analysis methodology development in support of the Savannah River Reactor PSA

International Nuclear Information System (INIS)

O'Kula, K.R.; Sharp, D.A.; Amos, C.N.; Wagner, K.C.; Bradley, D.R.

1992-01-01

A three-level Probabilistic Safety Assessment (PSA) of production reactor operation has been underway since 1985 at the US Department of Energy's Savannah River Site (SRS). The goals of this analysis are to: Analyze existing margins of safety provided by the heavy-water reactor (HWR) design challenged by postulated severe accidents; Compare measures of risk to the general public and onsite workers to guideline values, as well as to those posed by commercial reactor operation; and Develop the methodology and database necessary to prioritize improvements to engineering safety systems and components, operator training, and engineering projects that contribute significantly to improving plant safety. PSA technical staff from the Westinghouse Savannah River Company (WSRC) and Science Applications International Corporation (SAIC) have performed the assessment despite two obstacles: A variable baseline plant configuration and power level; and a lack of technically applicable code methodology to model the SRS reactor conditions. This paper discusses the detailed effort necessary to modify the requisite codes before accident analysis insights for the risk assessment were obtained

A methodology for supporting decisions on the establishment of protective measures after severe nuclear accidents. Final report

International Nuclear Information System (INIS)

Papazoglou, I.A.; Kollas, J.G.

1994-06-01

Full text: The objective of this report is to demonstrate the use of a methodology supporting decisions on protective measures following severe nuclear accidents. A multicriteria decision analysis approach is recommended where value tradeoffs are postponed until the very last stage of the decision process. Use of efficient frontiers is made to exclude all technically inferior solutions and present the decision maker with all non-dominated solutions. A choice among these solutions implies a value trade-off among the multiple criteria. An interactive computer package has been developed where the decision maker can choose a point on the efficient frontier in the consequence space and immediately see the alternative in the decision space resulting in the chosen consequences. The methodology is demonstrated through an application on the choice among possible protective measures in contaminated areas of the former USSR after the Chernobyl accident. Two distinct cases are considered: First a decision is to be made only on the basis of the level of soil contamination with Cs-137 and the total cost of the chosen protective policy; Next the decision is based on the geographic dimension of the contamination and the total cost. Three alternative countermeasure actions are considered for population segments living on soil contaminated at a certain level or in a specific geographic region: (a) relocation of the population; (b) improvement of the living conditions; and, (c) no countermeasures at all. This is the final deliverable of the CEC-CIS Joint Study Project 2, Task 5: Decision-Aiding-System for Establishing Intervention Levels, performed under Contracts COSU-CT91-0007 and COSU-CT92-0021 with the Commission of European Communities through CEPN. (author)

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

Directory of Open Access Journals (Sweden)

Zhou Yao

2018-01-01

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

Canister Storage Building (CSB) Design Basis Accident Analysis Documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

1999-09-09

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

Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

Energy Technology Data Exchange (ETDEWEB)

Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima, E-mail: dayanefs@ipen.br, E-mail: gdjian@ipen.br, E-mail: aclima@ipen.br [Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil)

2017-07-01

The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

Containment pressure analysis methodology during a LBLOCA with iteration between RELAP5 and COCOSYS

International Nuclear Information System (INIS)

Silva, Dayane Faria; Sabundjian, Gaianê; Souza, Ana Cecília Lima

2017-01-01

The pressure conditions inside the containment in the case of a Large Break Loss of Coolant Accident (LBLOCA) are more severe in the case of hot leg rupture, due to the large amount of mass and energy that is thrown from the break that lies just after the pressure vessel. This work presents a methodology of pressure analysis within the containment of a Brazilian PWR, Angra 2, with an iterative process between the code that simulates guillotine rupture - RELAP5 - and the COCOSYS code, which analyzes the containment pressure from the accident conditions. The results show that the iterative process between the codes allows the convergence of pressure data to a more realistic approach. (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-04-15

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

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

International Nuclear Information System (INIS)

Jung, Gunhyo; Jae, Moosung

2015-01-01

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

Review of accident analyses of RB experimental reactor

International Nuclear Information System (INIS)

Pesic, M.

2003-01-01

The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VINCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62; yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin) consisting of 2% enriched uranium metal and 80% enriched U0 2 , dispersed in aluminum matrix, have been available since 1962 and 1976, respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements, as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINCA Institute, an independent regulator)' body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety' Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed) to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given. (author)

Review of accident analyses of RB experimental reactor

Directory of Open Access Journals (Sweden)

Pešić Milan P.

2003-01-01

Full Text Available The RB reactor is a uranium fuel heavy water moderated critical assembly that has been put and kept in operation by the VTNCA Institute of Nuclear Sciences, Belgrade, Serbia and Montenegro, since April 1958. The first complete Safety Analysis Report of the RB reactor was prepared in 1961/62 yet, the first accident analysis had been made in late 1958 with the aim to examine a power transition and the total equivalent doses received by the staff during the reactivity accident that occurred on October 15, 1958. Since 1960, the RB reactor has been modified a few times. Beside the initial natural uranium metal fuel rods, new types of fuel (TVR-S types of Russian origin consisting of 2% enriched uranium metal and 80% enriched UO2 dispersed in aluminum matrix, have been available since 1962 and 1976 respectively. Modifications of the control and safety systems of the reactor were made occasionally. Special reactor cores were designed and constructed using all three types of fuel elements as well as the coupled fast-thermal ones. The Nuclear Safety Committee of the VINĆA Institute, an independent regulatory body, approved for usage all these modifications of the RB reactor on the basis of the Preliminary Safety Analysis Reports, which, beside proposed technical modifications and new regulation rules, included safety analyses of various possible accidents. A special attention was given (and a new safety methodology was proposed to thorough analyses of the design-based accidents related to the coupled fast-thermal cores that included central zones of the reactor filled by the fuel elements without any moderator. In this paper, an overview of some accidents, methodologies and computation tools used for the accident analyses of the RB reactor is given.

Pedestrian-Vehicle Accidents Reconstruction with PC-Crash®: Sensibility Analysis of Factors Variation

Energy Technology Data Exchange (ETDEWEB)

Martinez Gala, F.

2016-07-01

This paper describes the main findings of a study performed by INSIA-UPM about the improvement of the reconstruction process of real world vehicle-pedestrian accidents using PC-Crash® software, aimed to develop a software tool for the estimation of the variability of the collision speed due to the lack of real values of some parameters required during the reconstruction task. The methodology has been based on a sensibility analysis of the factors variation. A total of 9 factors have been analyzed with the objective of identifying which ones were significant. Four of them (pedestrian height, collision angle, hood height and pedestrian-road friction coefficient) were significant and were included in a full factorial experiment with the collision speed as an additional factor in order to obtain a regression model with up to third level interactions. Two different factorial experiments with the same structure have been performed because of pedestrian gender differences. The tool has been created as a collision speed predictor based on the regression models obtained, using the 4 significant factors and the projection distance measured or estimated in the accident site. The tool has been used on the analysis of real-world reconstructed accidents occurred in the city of Madrid (Spain). The results have been adequate in most cases with less than 10% of deviation between the predicted speed and the one estimated in the reconstructions. (Author)

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

Science.gov (United States)

Arya Dewanto, Y.; Faturachman, D.

2018-03-01

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

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

International Nuclear Information System (INIS)

Svenson, Ola

2000-02-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2000-02-01

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

Accident Sequence Evaluation Program: Human reliability analysis procedure

Energy Technology Data Exchange (ETDEWEB)

Swain, A.D.

1987-02-01

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

Macro Data Analysis of Traffic Accidents in Indonesia

Directory of Open Access Journals (Sweden)

Annisa Jusuf

2017-04-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Severe accidents at nuclear power plants. Their risk assessment and accident management

International Nuclear Information System (INIS)

Abe, Kiyoharu.

1995-05-01

This document is to explain the severe accident issues. Severe Accidents are defined as accidents which are far beyond the design basis and result in severe damage of the core. Accidents at Three Mild Island in USA and at Chernobyl in former Soviet Union are examples of severe accidents. The causes and progressions of the accidents as well as the actions taken are described. Probabilistic Safety Assessment (PSA) is a method to estimate the risk of severe accidents at nuclear reactors. The methodology for PSA is briefly described and current status on its application to safety related issues is introduced. The acceptability of the risks which inherently accompany every technology is then discussed. Finally, provision of accident management in Japan is introduced, including the description of accident management measures proposed for BWRs and PWRs. (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).

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

International Nuclear Information System (INIS)

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

1990-01-01

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

Safety analysis methodology for Chinshan nuclear power plant spent fuel pool under Fukushima-like accident condition

Energy Technology Data Exchange (ETDEWEB)

Lin, Hao-Tzu [Institute of Nuclear Energy Research, Taoyuan, Taiwan (China). Research Atomic Energy Council; Li, Wan-Yun; Wang, Jong-Rong; Tseng, Yung-Shin; Chen, Hsiung-Chih; Shih, Chunkuan; Chen, Shao-Wen [National Tsing Hua Univ., HsinChu, Taiwan (China). Inst. of Nuclear Engineering and Science

2017-03-15

Chinshan nuclear power plant (NPP), a BWR/4 plant, is the first NPP in Taiwan. After Fukushima NPP disaster occurred, there is more concern for the safety of NPPs in Taiwan. Therefore, in order to estimate the safety of Chinshan NPP spent fuel pool (SFP), by using TRACE, MELCOR, CFD, and FRAPTRAN codes, INER (Institute of Nuclear Energy Research, Atomic Energy Council, R.O.C.) performed the safety analysis of Chinshan NPP SFP. There were two main steps in this research. The first step was the establishment of Chinshan NPP SFP models. And the transient analysis under the SFP cooling system failure condition (Fukushima-like accident) was performed. In addition, the sensitive study of the time point for water spray was also performed. The next step was the fuel rod performance analysis by using FRAPTRAN and TRACE's results. Finally, the animation model of Chinshan NPP SFP was presented by using the animation function of SNAP with MELCOR analysis results.

[Spatial analysis of road traffic accidents with fatalities in Spain, 2008-2011].

Science.gov (United States)

Gómez-Barroso, Diana; López-Cuadrado, Teresa; Llácer, Alicia; Palmera Suárez, Rocío; Fernández-Cuenca, Rafael

2015-09-01

To estimate the areas of greatest density of road traffic accidents with fatalities at 24 hours per km(2)/year in Spain from 2008 to 2011, using a geographic information system. Accidents were geocodified using the road and kilometer points where they occurred. The average nearest neighbor was calculated to detect possible clusters and to obtain the bandwidth for kernel density estimation. A total of 4775 accidents were analyzed, of which 73.3% occurred on conventional roads. The estimated average distance between accidents was 1,242 meters, and the average expected distance was 10,738 meters. The nearest neighbor index was 0.11, indicating that there were aggregations of accidents in space. A map showing the kernel density was obtained with a resolution of 1 km(2), which identified the areas of highest density. This methodology allowed a better approximation to locating accident risks by taking into account kilometer points. The map shows areas where there was a greater density of accidents. This could be an advantage in decision-making by the relevant authorities. Copyright © 2014 SESPAS. Published by Elsevier Espana. All rights reserved.

Analysis of construction accidents in Spain, 2003-2008.

Science.gov (United States)

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

2012-12-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-10-01

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

The Application of Best Estimate and Uncertainty Analysis Methodology to Large LOCA Power Pulse in a CANDU 6 Reactor

International Nuclear Information System (INIS)

Abdul-Razzak, A.; Zhang, J.; Sills, H.E.; Flatt, L.; Jenkins, D.; Wallace, D.J.; Popov, N.

2002-01-01

The paper describes briefly a best estimate plus uncertainty analysis (BE+UA) methodology and presents its proto-typing application to the power pulse phase of a limiting large Loss-of-Coolant Accident (LOCA) for a CANDU 6 reactor fuelled with CANFLEX R fuel. The methodology is consistent with and builds on world practice. The analysis is divided into two phases to focus on the dominant parameters for each phase and to allow for the consideration of all identified highly ranked parameters in the statistical analysis and response surface fits for margin parameters. The objective of this analysis is to quantify improvements in predicted safety margins under best estimate conditions. (authors)

Radiological accidents: analysis of the information disseminated by media and public acceptance of nuclear technology

International Nuclear Information System (INIS)

Delgado, Jose Ubiratan; Tauhata, Luiz; Garcia, Marcia Maria

1995-01-01

A methodology to treat quantitatively information by Media concerning a nuclear or a radiological accident is presented. It allows us to classify information according to the amount, importance and way of showing, into one indicator, named Information Equivalent. This establishes a procedure for analysis of released information and includes: number of head-lines, illustrations, printed lines, editorials, authorities quoted and so on. Interpretation becomes easier when the evolution and statistical trend of this indicator is observed. The application to evaluate the dissemination of the accident which took place in 1987 in Goiania, Brazil, was satisfactory and allowed us to purpose a model. This will aid the planning, the decision making process and it will improve relationships between technical staff and media during the emergency. (author). 5 refs., 4 figs., 3 tabs

Analysis of hypothetical LMFBR whole-core accidents in the USA

International Nuclear Information System (INIS)

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

1978-01-01

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

Accident Sequence Precursor Analysis for SGTR by Using Dynamic PSA Approach

International Nuclear Information System (INIS)

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

2016-01-01

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

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

Development of the criticality accident analysis code, AGNES

International Nuclear Information System (INIS)

Nakajima, Ken

1989-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

PIEPHO, M.G.

1999-10-20

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

A methodology for optimisation of countermeasures for animal products after a nuclear accident and its application

International Nuclear Information System (INIS)

Hwang, Won Tae; Cho, Gyuseong; Han, Moon Hee

1999-01-01

A methodology for the optimisation of the countermeasures associated with the contamination of animal products was designed based on cost-benefit analysis. Results are discussed for the hypothetical deposition of radionuclides on 15 August, when pastures are fully developed in Korean agricultural conditions. A dynamic food chain model, DYNACON, was used to evaluate the effectiveness of the countermeasures for reducing the ingestion dose. The countermeasures considered were: (1) a ban on food consumption; and (2) the substitution of clean fodder. These are effective in reducing the ingestion dose as well as simple and easy to carry out in the first year after deposition. The net benefit of the countermeasures was quantitatively estimated in terms of avertable doses and monetary costs. The benefit depends on a variety of factors, such as radionuclide concentrations on the ground, starting time and duration of the countermeasures. It is obvious that a fast reaction after deposition is important in maximising the cost effectiveness of the countermeasures. In most cases, the substitution of clean fodder is more cost effective than a ban on food consumption. The methodology used in this study may serve as a basis for rapid decision-making on the introduction of countermeasures relating to the contamination of animal products after a nuclear accident

Systematic approach for assessment of accident risks in chemical and nuclear processing; Abordagem sistematica para avaliacao de riscos de acidentes em instalacoes de processamento quimico e nuclear

Energy Technology Data Exchange (ETDEWEB)

Senne, Junior, Murillo

2003-07-15

The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of systematizing

Systematic approach for assessment of accident risks in chemical and nuclear processing; Abordagem sistematica para avaliacao de riscos de acidentes em instalacoes de processamento quimico e nuclear

Energy Technology Data Exchange (ETDEWEB)

Senne Junior, Murillo

2003-07-15

The industrial accidents which occurred in the last years, particularly in the 80's, contributed a significant way to draw the attention of the government, industry and the society as a whole to the mechanisms for preventing events that could affect people's safety and the environment quality. Techniques and methods extensively used the nuclear, aeronautic and war industries so far were adapted to performing analysis and evaluation of the risks associated to other industrial activities, especially in the petroleum, chemistry and petrochemical areas. The risk analysis in industrial facilities is carried out through the evaluation of the probability or frequency of the accidents and their consequences. However, no systematized methodology that could supply the tools for identifying possible accidents likely to take place in an installation is available in the literature. Neither existing are methodologies for the identification of the models for evaluation of the accidents' consequences nor for the selection of the available techniques for qualitative or quantitative analysis of the possibility of occurrence of the accident being focused. The objective of this work is to develop and implement a methodology for identification of the risks of accidents in chemical and nuclear processing facilities as well as for the evaluation of their consequences on persons. For the development of the methodology, the main possible accidents that could occur in such installations were identified and the qualitative and quantitative techniques available for the identification of the risks and for the evaluation of the consequences of each identified accidents were selected. The use of the methodology was illustrated by applying it in two case examples adapted from the literature, involving accidents with inflammable, explosives, and radioactive materials. The computer code MRA - Methodology for Risk Assessment was developed using DELPHI, version 5.0, with the purpose of

Accident sequence analysis of human-computer interface design

International Nuclear Information System (INIS)

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

2000-01-01

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

Radiological accidents: methodologies of radio nuclides dis incorporation

International Nuclear Information System (INIS)

Jimenez F, E. A.; Paredes G, L.; Cortes, A.

2014-08-01

Derived of the radioactive or nuclear material management, exists the risk that accidents can happen where people cases are presented with internal radioactive contamination, who will receive specialized medical care to accelerate the radioactive dis incorporation with the purpose of diminishing the absorbed dose and the associate biological effects. In this work treatments of radioactive dis incorporation were identified, in function of the radionuclide, radiation type, radioactive half life, biological half life, critical organ, ingestion duct and patient type. The factor time is decisive for the effectiveness of the selected treatment in the blockade stage (before the accident) or dis incorporation (after the accident); this factor is related with the radioactive and biological half lives. So to achieve dis incorporation efficiencies of more to 70%, the patient clinical treatment will begin before the first third of the biological half life of the radionuclide that generated the internal contamination. (Author)

Case for integral core-disruptive accident analysis

International Nuclear Information System (INIS)

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

1985-01-01

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

The prevention of radiological accidents (how to avoid or minimize potential exposures)

International Nuclear Information System (INIS)

Touzet, Rodolfo E.

2006-01-01

A detailed analysis of 7 major accidents occurred in radiotherapy services in different countries is performed. Then a generic analysis of the causes is realized and finally the methodology used to prevent them effectively is described [es

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

International Nuclear Information System (INIS)

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

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2015-05-15

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

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.

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

International Nuclear Information System (INIS)

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

2014-01-01

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

A study on the pressurized water reactor (PWR) containment response analysis methodologies for postulated severe accident

International Nuclear Information System (INIS)

Ahn, Kwang Il

1992-02-01

The present study contains two major parts: one is the treatment of uncertainties involved in the current APET and the other is the importance analysis of the APET uncertainty inputs. A clear disadvantage of the expert opinion polling process approach for uncertainty analysis of the current probabilistic risk assessment (PRA) is that the sufficient robustness in the final results may not be attained against the ambiguity of the information upon which the experts base their judgement or the judgmental uncertainty arising under various imprecise and incomplete information. For the treatment of such type of uncertainty, a new approach based on fuzzy set theory is proposed. Then its potential use to the uncertainty analysis of the current PRA is proved through an analysis of accident progression event tree (APET). As a product, a formal procedure with computational algorithms suitable for application of the fuzzy set theory to the APET analysis is provided. Comparing with the uncertainty analysis results obtained by the statistical approach currently used in PRA, the present approach has several major advantages: Firstly, it greatly enhances the robustness in the final results of APET uncertainty analysis by modeling the judgmental uncertainty that arises in the probabilistic quantification of APET top events. Secondly, the modeling of APET uncertainty analysis is far more convenient because of the nonprobabilistic features of fuzzy probabilities used for uncertainty quantification of the APET top events. Thirdly, the APET model can easily be operated by means of a well defined formal propagation logic of fuzzy set theory without going through a tedious sampling procedure. Finally, the fuzzy outcomes provide at least as much information as the existing methods based on the statistical approach. Thus, the present approach can be used as a valuable alternative approach to uncertainty analysis used in the current PRA. Two importance measures for the importance analysis of

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

International Nuclear Information System (INIS)

Song, Y.M.; Kim, D.H.; Nijhawan, Sunil

2015-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

Song, Y.M.; Kim, D.H. [KAERI, Daejeon (Korea, Republic of); Nijhawan, Sunil [Prolet Inc. 98 Burbank Drive, Toronto (Canada)

2015-05-15

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

Hazards and accident analyses, an integrated approach, for the Plutonium Facility at Los Alamos National Laboratory

International Nuclear Information System (INIS)

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

1995-01-01

This paper describes an integrated approach to perform hazards and accident analyses for the Plutonium Facility at Los Alamos National Laboratory. A comprehensive hazards analysis methodology was developed that extends the scope of the preliminary/process hazard analysis methods described in the AIChE Guidelines for Hazard Evaluations. Results fro the semi-quantitative approach constitute a full spectrum of hazards. For each accident scenario identified, there is a binning assigned for the event likelihood and consequence severity. In addition, each accident scenario is analyzed for four possible sectors (workers, on-site personnel, public, and environment). A screening process was developed to link the hazard analysis to the accident analysis. Specifically the 840 accident scenarios were screened down to about 15 accident scenarios for a more through deterministic analysis to define the operational safety envelope. The mechanics of the screening process in the selection of final scenarios for each representative accident category, i.e., fire, explosion, criticality, and spill, is described

Assessment of PASS Effectiveness under Severe Accidents in Nuclear Power Plants

International Nuclear Information System (INIS)

Choi, Yu Jung; Lee, Sung Bok; Kim, Hyeong Taek; Lee, Jin Yong

2008-01-01

Following the accident at Three Mile Island Unit 2 (TMI-2) on March 28, 1979, the USNRC formed a lessons-learned Task Force to identify and evaluate safety concerns originating with the TMI-2 accident. NUREG-0578 documented the results of the task force effort. One of the recommendations of the task force was for licensees to upgrade the capability to obtain samples from the reactor coolant system and containment atmosphere under high radioactivity conditions and to provide the capability for chemical and spectral analyses of high-level samples on site. NUREG-0737 contained the details of the TMI recommendations that were to be implemented by the licensees. Additional criteria for post accident sampling system(PASS) were issued by Regulatory Guide 1.97. As the results, PASS has been installed on nuclear power plants(NPPs) in Korea as well as United States. However, significant improvements have been achieved since the TMI-2 accident in the areas of understanding risks associated with nuclear plant operations and developing better strategies for managing the response to potential severe accidents at NPPs. Thus, the requirements for PASS have been re-evaluated in some reports. According to the reports, the samples and measurements from PASS do not contribute significantly to emergency management response to severe accidents due to the long analyzing time, 3 hours. Hence, this paper focused on the development of the quantitative analysis methodology to analyze the sequence of the severe accident in Yonggwang nuclear power plants (YGN) and presented the results of the analysis according to the developed methodology

Study of possibility using LANL PSA-methodology for accident probability RBMK researches

International Nuclear Information System (INIS)

Petrin, S.V.; Yuferev, V.Y.; Zlobin, A.M.

1995-01-01

The reactor facility probabilistic safety analysis methodologies are considered which are used at U.S. LANL and RF NIKIET. The methodologies are compared in order to reveal their similarity and differences, determine possibilities of using the LANL technique for RBMK type reactor safety analysis. It is found that at the PSA-1 level the methodologies practically do not differ. At LANL the PHA, HAZOP hazards analysis methods are used for more complete specification of the accounted initial event list which can be also useful at performance of PSA for RBMK. Exchange of information regarding the methodology of detection of dependent faults and consideration of human factor impact on reactor safety is reasonable. It is accepted as useful to make a comparative study result analysis for test problems or PSA fragments using various computer programs employed at NIKIET and LANL

The DOE technology development programme on severe accident management

International Nuclear Information System (INIS)

Neuhold, R.J.; Moore, R.A.; Theofanous, T.G.

1998-01-01

The US Department of Energy (DOE) is sponsoring a programme in technology development aimed at resolving the technical issues in severe accident management strategies for advanced and evolutionary light water reactors (LWRs). The key objective of this effort is to achieve a robust defense-in-depth at the interface between prevention and mitigation of severe accidents. The approach taken towards this goal is based on the Risk Oriented Accident Analysis Methodology (ROAAM). Applications of ROAAM to the severe accident management strategy for the US AP600 advanced LWR have been effective both in enhancing the design and in achieving acceptance of the conclusions and base technology developed in the course of the work. This paper presents an overview of that effort and its key technical elements

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

International Nuclear Information System (INIS)

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

2011-06-01

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

Application of NUREG-1150 methods and results to accident management

International Nuclear Information System (INIS)

Dingman, S.E.; Sype, T.T.; Camp, A.L.

1990-01-01

The risk from five nuclear power plants was examined during the NUREG-1150 program. When the analyses of the plants were complete, an effort was undertaken to examine the implications of NUREG-1150 for accident management initiatives. The framework provided by the NUREG-1150 analysis presented a means within which current accident management strategies could be evaluated and future accident management strategies could be developed and assessed. Five separate but interrelated phases of risk management were considered: (1) prevention of accident initiators, (2) prevention of core damage, (3) implementation of an effective emergency response, (4) prevention of vessel breach and mitigation of radionuclide releases from the reactor coolant system, and (5) retention of fission products in the containment and other surrounding buildings. A risk-based methodology was developed to identify and evaluate risk management options for each of these five phases. The methodology was demonstrated through quantitative examples for the first two phases of risk management listed above. In addition, the reduction in risk for several currently implemented risk management strategies at operating plants was quantified

Analysis of Public Perception on Radiation: with One Year after Fukushima Nuclear Accident

Energy Technology Data Exchange (ETDEWEB)

Park, Bang Ju [Korean Science Reporters Association, Seoul (Korea, Republic of)

2012-03-15

A year has passed since the nuclear power plant accident in Fukushima on March 11, 2011, and a survey for public perception on radiation by Korean people has been made. The methodological design was based on a quantitative survey and a frequency analysis was done. The analysis objects were survey papers (n=2,754pcs) answered by random ordinary citizens chosen from all over the country. The questionnaires, and study tool, were directly distributed and collected. A total of 40 questionnaires using a coefficient of Cronbach's {alpha} per each area was 'self perception of radiation' (0.620), 'radiation risk' (0.830), 'benefit from radiation' (0.781), 'radiation controlled' (0.685), 'informative source of radiation' (0.831), 'influence degree from Fukushima accident' (0.763), showing rather high score from all areas. As the result of the questionnaires, the knowledge of radiation concept was 69.50 out of 100 points, which shows a rather significant difference from the result of 'know well about radiation' (53.7%) and 'just know about radiation' (37.40%). According to the survey, one of the main reasons why radiation seems risky was that once exposed to radiation, it may not have negative impacts presently but, the next generation could see negative impacts (66.1%). About 41% of our respondents showed a negative position against the government's report on radiation while 39.5% of respondents said that we should stop running nuclear power in light of Fukushima nuclear power plant accident. This study was done for the first time by Korean people's public perception on radiation after the Fukushima nuclear power plant accident. We expect this might have significant contributions to the establishment of the government's policy on radiation.

Analysis of Public Perception on Radiation: with One Year after Fukushima Nuclear Accident

International Nuclear Information System (INIS)

Park, Bang Ju

2012-01-01

A year has passed since the nuclear power plant accident in Fukushima on March 11, 2011, and a survey for public perception on radiation by Korean people has been made. The methodological design was based on a quantitative survey and a frequency analysis was done. The analysis objects were survey papers (n=2,754pcs) answered by random ordinary citizens chosen from all over the country. The questionnaires, and study tool, were directly distributed and collected. A total of 40 questionnaires using a coefficient of Cronbach's α per each area was 'self perception of radiation' (0.620), 'radiation risk' (0.830), 'benefit from radiation' (0.781), 'radiation controlled' (0.685), 'informative source of radiation' (0.831), 'influence degree from Fukushima accident' (0.763), showing rather high score from all areas. As the result of the questionnaires, the knowledge of radiation concept was 69.50 out of 100 points, which shows a rather significant difference from the result of 'know well about radiation' (53.7%) and 'just know about radiation' (37.40%). According to the survey, one of the main reasons why radiation seems risky was that once exposed to radiation, it may not have negative impacts presently but, the next generation could see negative impacts (66.1%). About 41% of our respondents showed a negative position against the government's report on radiation while 39.5% of respondents said that we should stop running nuclear power in light of Fukushima nuclear power plant accident. This study was done for the first time by Korean people's public perception on radiation after the Fukushima nuclear power plant accident. We expect this might have significant contributions to the establishment of the government's policy on radiation.

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)

Review of accident analyses performed at Mochovce NPP

International Nuclear Information System (INIS)

Siko, D.

2000-01-01

In this paper the review of accident analysis performed in NPP Mochovce V-1 is presented. The scope of these safety measures was defined and development in the T SSM for NPP Mochovce Nuclear Safety Improvements Report' issued in July 1995. The main objectives of these safety measures were the followings: (a) to establish the criteria for selection and classification of accidental events, as well as defining the list of initiating events to be analysed. Accident classification to the individual groups must be performed in accordance with RG 1.70 and IAEA recommendations 'Guidelines for Accidental Analysis of WWER NPP' (IAEA-EBR-WWER-01) to select boundary cases to be calculated from the scope of initiating events; (b ) to elaborate the accident analysis methodology that also includes acceptance criteria for their result evaluation, initial and boundary conditions, assumption related with the application of the single failure criteria, requirements on the analysis quality, used computer codes, as well as NPP models and input data for the accident analysis; (c) to perform the accident analysis for the Pre-operational Safety Report (POSAR); (d) to provide a synthetic report addressing the validity range of codes models and correlations, the assessment against relevant tests results, the evidence of the user qualification, the modernisation and nodding scheme for the plant and the justification of used computer codes. Analyses results showed that all acceptance criteria were met with satisfactory margin and design of the NPP Mochovce is accurate. (author)

Historical survey of the qualifying process of Furnas calculus methodology in the areas of rods, neutronics, thermohydraulic accidents and transients

International Nuclear Information System (INIS)

Conti, C.F.S.; Silva Galetti, M.R. da.

1990-02-01

As Furnas intends to assume in the future the responsibility of performing Safety Analyses associated to Reload and Operation questions to Angra 1, it was figured out the necessity of qualifying its methodology by CNEN. The Methodology Qualification Process is based on guidelines proposed by CNEN at NT-DR-N o 02/87, where it was divided in four steps. This Technical Note aims to present the follow up of FURNAS Methodology Qualification Process and to bring it up to date in the areas of Core Physics (Neutronics), Core Thermal-Hydraulics, Fuel Rod Behaviour, Transient and Large Break Loss of Coolant Accident Analyses (LBLOCA). (author)

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

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

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

Methodological approaches to analysis of agricultural countermeasures on radioactive contaminated areas: Estimation of effectiveness and comparison of different alternatives

DEFF Research Database (Denmark)

Yatsalo, B.I.; Hedemann Jensen, P.; Alexakhin, R.M.

1997-01-01

Methodological aspects of countermeasure analysis in the long-term period after a nuclear accident are discussed for agriculture countermeasures for illustrative purposes. The estimates of effectiveness fbr specific countermeasures as well as methods of justified action levels assessments...... and comparison of different alternatives (countermeasures) based on the use of several criteria are considered....

Analysis of accident sequences and source terms at waste treatment and storage facilities for waste generated by U.S. Department of Energy Waste Management Operations, Volume 1: Sections 1-9

International Nuclear Information System (INIS)

Mueller, C.; Nabelssi, B.; Roglans-Ribas, J.

1995-04-01

This report documents the methodology, computational framework, and results of facility accident analyses performed for the U.S. Department of Energy (DOE) Waste Management Programmatic Environmental Impact Statement (WM PEIS). The accident sequences potentially important to human health risk are specified, their frequencies are assessed, and the resultant radiological and chemical source terms are evaluated. A personal computer-based computational framework and database have been developed that provide these results as input to the WM PEIS for calculation of human health risk impacts. The methodology is in compliance with the most recent guidance from DOE. It considers the spectrum of accident sequences that could occur in activities covered by the WM PEIS and uses a graded approach emphasizing the risk-dominant scenarios to facilitate discrimination among the various WM PEIS alternatives. Although it allows reasonable estimates of the risk impacts associated with each alternative, the main goal of the accident analysis methodology is to allow reliable estimates of the relative risks among the alternatives. The WM PEIS addresses management of five waste streams in the DOE complex: low-level waste (LLW), hazardous waste (HW), high-level waste (HLW), low-level mixed waste (LLMW), and transuranic waste (TRUW). Currently projected waste generation rates, storage inventories, and treatment process throughputs have been calculated for each of the waste streams. This report summarizes the accident analyses and aggregates the key results for each of the waste streams. Source terms are estimated and results are presented for each of the major DOE sites and facilities by WM PEIS alternative for each waste stream. The appendices identify the potential atmospheric release of each toxic chemical or radionuclide for each accident scenario studied. They also provide discussion of specific accident analysis data and guidance used or consulted in this report

Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

International Nuclear Information System (INIS)

Park, J. Y.; Park, Y. W.; Park, H.G.

2016-01-01

This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly

Case Study on Influence Factor Trend Analysis of the Accidents and Events of Nuclear Power Plants by applying Nuclear Safety Culture Framework

Energy Technology Data Exchange (ETDEWEB)

Park, J. Y.; Park, Y. W.; Park, H.G. [KAIST, Daejeon (Korea, Republic of)

2016-05-15

This study 1) established the standard based on frameworks of safety culture principles that show safety culture promotion goals, 2) analyzed the linkages with the frameworks that were established by analyzing each incident cause and weak point from selected 268 cases(rating over INES grade 1) among 4,088 cases (as of April 1, 2015). The 4,088 cases were selected as a result of database analysis from 702 accidents recorded in accident and rating evaluation reports that were published in the National Nuclear Safety Commission and overseas IRS (International Reporting System for operating Experience), and 3) finally conducted a trend analysis studies with these comprehensive results. From the investigations, followings were concluded. 1) In order to analyze the safety culture, analysis methodology is required. 2) Analytical methodology for building sustainable safety culture promoting a virtuous cycle system was developed 3) Among variety of process input data, 970 domestic and overseas incidents were selected as targets and 502 accidents were classified as safety culture related events by utilizing screen filter of IAEA GS-G-3.5 Appendix I and Framework (Nuclear Safety Culture Base Frame) developed by BEES, Inc. for safety culture analysis method. 4) As a result, complex safety culture influence factors for the one reason which was difficult to separate by conventional methods was able to be analyzed. 5) The cumulative data through the system was results of virtuous trend analysis rather than temporary results. Thus, it could be unique cultural factors of the domestic industry and could derive trend differences for domestic safety culture factors accordingly.

Current status of accident analysis for Korean HCCR TBS

Energy Technology Data Exchange (ETDEWEB)

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

2014-10-15

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

Current status of accident analysis for Korean HCCR TBS

International Nuclear Information System (INIS)

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

2014-01-01

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

Methods for air cleaning system design and accident analysis

International Nuclear Information System (INIS)

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

1987-01-01

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

Problems of probabilistic safety assessment after Fukushima Daiichi nuclear power plant accident

International Nuclear Information System (INIS)

Sugiyama, Naoki

2011-01-01

Probabilistic safety assessment (PSA) methodology to assure nuclear safety is had great expectations of lessons learned from Fukushima Daiichi nuclear power plant (NPP) accident and on the other hand this accident made actualized technical problems of PSA. Effectiveness of current PSA methodology for risk assessment was confirmed by comparing the accident development with accident scenario of PSA and equipment failure rate. From a viewpoint of nuclear safety objective and defense in depth approach of IAEA, technical problems of PSA were (1) extension of PSA for spent fuel pool and waste disposal system as well as level 3PSA for broader environmental contamination and (2) overlapping of accident scenario of plural unit site, balance of high quality plant management and preceding negation, treatment of uncertainty of external events, severe accident measure and human reliability analysis and reflection of disaster prevention capability to level 3PSA. In order to upgrade PSA technology, six proposals were described for nuclear safety and defense in depth, comprehensive evaluation scope and catch-up of latest technology, necessity of strategic preparation of PSA standard, human resources fostering and risk communication. (T. Tanaka)

Analysis of traffic accidents in Romania, 2009.

Science.gov (United States)

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

2012-01-01

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

Cold Vacuum Drying facility design basis accident analysis documentation

International Nuclear Information System (INIS)

CROWE, R.D.

2000-01-01

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

Cold Vacuum Drying facility design basis accident analysis documentation

Energy Technology Data Exchange (ETDEWEB)

CROWE, R.D.

2000-08-08

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

Accidents in nuclear facilities: classification, incidence and impact

International Nuclear Information System (INIS)

Galicia A, J.; Paredes G, L. C.

2012-10-01

A general analysis of the 146 accidents reported officially in nuclear facilities from 1945 to 2012 is presented, among them some took place in: power or research nuclear reactors, critical and subcritical nuclear assemblies, handling of nuclear materials inside laboratories belonging to institutes or universities, in radiochemistry industrial plants and nuclear fuel factories. In form graph the incidence of these accidents is illustrated classified for; category, decades, geographical localization, country classification before the OECD, failure type, and the immediate or later victims. On the other hand, the main learned lessons of the nuclear accidents of Three Mile Island, Chernobyl and Fukushima are stood out, among those that highlight; the human factors, the necessity of designs more innovative and major technology for the operation, control and surveillance of the nuclear facilities, to increase the criterions of nuclear, radiological and physics safety applied to these facilities, the necessity to carry out probabilistic analysis of safety more detailed for cases of not very probable accidents and their impact, to revalue the selection criterions of the sites for nuclear locations, the methodology of post-accident sites recovery and major instrumentation for parameters evaluation and the radiological monitoring among others. (Author)

Outcomes from the regional Co-operation in the Area of the Safety Analysis Methodology

International Nuclear Information System (INIS)

D'Auria, F.; Mavko, B.; Prosek, A.; Debrecin, N.; Bajs, T.

2000-01-01

International Atomic Energy Agency (IAEA) carried out the Co-ordinated Research Program (CRP) ON V alidation of Accident and Safety Analysis Methodology'' in the period between 1995 and 1998. Three areas of interest identified by the participants referred to the pressurised water reactors of Western and Eastern type (PWR and WWER type). The specific areas of attention were: system behaviour of the primary and secondary loops (PS area), the containment response (CO area) and the severe accidents (SA area). During the CRP it became clear that the technology advancements, the available tools (i.e. codes) and the experimental databases in the above areas are quite different. At the conclusion of the CRP, all objectives of the program have been reached. This paper presents the summary of the regional co-operation in this framework. The CRP activities focused on the codes and expertise available at the participating organisations. This overview therefore summarises their experience related to the state-of-the-art in the field of computational accident analysis. In addition, the paper proposes the recommendations for future activities related to the code usage, the user effects and code development. In pursuing of these goals special attention is given to the importance of the international co-operation. (author)

Causal factors in accidents of high-speed craft and conventional ocean-going vessels

International Nuclear Information System (INIS)

Antao, Pedro; Guedes Soares, C.

2008-01-01

An analysis of 40 ocean-going commercial vessel accidents is compared with the study of a similar number of high-speed crafts (HSCs) accidents, using in both cases a methodology that highlights the sequence of events leading to the accident and identifies the associated latent or causal factors. The main objective of this study was to identify and understand the difference in the pattern of causal factors associated with HSC accidents, as compared with the more traditional ocean-going ships. From the analysis one can see that the HSC accidents are mainly related to bridge personnel and operations, where the human element is the key factor identified as being responsible for the majority of the accidents. When compared with ocean-going commercial vessels, it is clear that navigational equipment and procedures have a larger preponderance in terms of the occurrence of accidents of HSC and particular attention should be given to these issues

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.

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

Risk-assessment methodology for fast breeder reactors

International Nuclear Information System (INIS)

Ott, K.O.

1976-04-01

The methods applied or proposed for risk assessment of nuclear reactors are reviewed, particularly with respect to their applicability for risk assessment of future commercial fast breeder reactors. All methods are based on the calculation of accident consequences for relatively few accident scenarios. The role and general impact of uncertainties in fast-reactor accident analysis are discussed. The discussion shows the need for improvement of the methodology. A generalized and improved risk-assessment methodology is outlined and proposed (accident-spectra-progression approach). The generalization consists primarily of an explicit treatment of uncertainties throughout the accident progression. The results of this method are obtained in form of consequence distributions. The width and shape of the distributions depend in part on the superposition of the uncertainties. The first moment of the consequence distribution gives an improved prediction of the ''average'' consequence. The higher-consequence moments can be used for consideration of risk aversion. The assessment of the risk of one or a certain number of nuclear reactors can only provide an ''isolated'' risk assessment. The general problem of safety risk assessment and its relation to public acceptance of certain modes of power production is a much broader problem area, which is also discussed

Evaluation of the status of science and technology concerning the performance and interpretation of accident analyses using analysis simulators; Ermittlung des Standes von Wissenschaft und Technik bei der Durchfuehrung und Bewertung von Stoerfallanalysen und der Verwendung von Analysesimulatoren

Energy Technology Data Exchange (ETDEWEB)

Pointner, Winfried; Cuesta Morales, Alejandra; Koppers, Vera; Kowalik, Michael; Broecker, Annette; Li, Tianqi; Mayer, Gerhard; Palazzo, Simone; Perin, Yann

2017-03-15

The aim of the project is to develop the methodology to perform short-term accident analyses and to clarify scientific-technical questions concerning the reactor and behavior in case of an accident. For short-term reactions it is necessary to provide appropriate analysis tools data and results for the plant specific behavior (reactor core, cooling circuits, containment) for abnormal operation and accidents (design-based and beyond-design accidents). The qualification of the data base and severe accident analyses is part of the project. The analysis simulators were actualized to the most recent version of ATHLET 3.1A.

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

CANDU safety analysis system establishment; development of trip coverage and multi-dimensional hydrogen analysis methodology

Energy Technology Data Exchange (ETDEWEB)

Choi, Jong Ho; Ohn, M. Y.; Cho, C. H. [KOPEC, Taejon (Korea)

2002-03-01

The trip coverage analysis model requires the geometry network for primary and secondary circuit as well as the plant control system to simulate all the possible plant operating conditions throughout the plant life. The model was validated for the power maneuvering and the Wolsong 4 commissioning test. The trip coverage map was produced for the large break loss of coolant accident and the complete loss of class IV power event. The reliable multi-dimensional hydrogen analysis requires the high capability for thermal hydraulic modelling. To acquire such a basic capability and verify the applicability of GOTHIC code, the assessment of heat transfer model, hydrogen mixing and combustion model was performed. Also, the assessment methodology for flame acceleration and deflagration-to-detonation transition is established. 22 refs., 120 figs., 31 tabs. (Author)

A Risk Analysis Methodology to Address Human and Organizational Factors in Offshore Drilling Safety: With an Emphasis on Negative Pressure Test

Science.gov (United States)

Tabibzadeh, Maryam

According to the final Presidential National Commission report on the BP Deepwater Horizon (DWH) blowout, there is need to "integrate more sophisticated risk assessment and risk management practices" in the oil industry. Reviewing the literature of the offshore drilling industry indicates that most of the developed risk analysis methodologies do not fully and more importantly, systematically address the contribution of Human and Organizational Factors (HOFs) in accident causation. This is while results of a comprehensive study, from 1988 to 2005, of more than 600 well-documented major failures in offshore structures show that approximately 80% of those failures were due to HOFs. In addition, lack of safety culture, as an issue related to HOFs, have been identified as a common contributing cause of many accidents in this industry. This dissertation introduces an integrated risk analysis methodology to systematically assess the critical role of human and organizational factors in offshore drilling safety. The proposed methodology in this research focuses on a specific procedure called Negative Pressure Test (NPT), as the primary method to ascertain well integrity during offshore drilling, and analyzes the contributing causes of misinterpreting such a critical test. In addition, the case study of the BP Deepwater Horizon accident and their conducted NPT is discussed. The risk analysis methodology in this dissertation consists of three different approaches and their integration constitutes the big picture of my whole methodology. The first approach is the comparative analysis of a "standard" NPT, which is proposed by the author, with the test conducted by the DWH crew. This analysis contributes to identifying the involved discrepancies between the two test procedures. The second approach is a conceptual risk assessment framework to analyze the causal factors of the identified mismatches in the previous step, as the main contributors of negative pressure test

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

Energy Technology Data Exchange (ETDEWEB)

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

1963-02-15

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

Historical analysis of US pipeline accidents triggered by natural hazards

Science.gov (United States)

Girgin, Serkan; Krausmann, Elisabeth

2015-04-01

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

Assessment of ISLOCA risk: Methodology and application to a Westinghouse four-loop ice condenser plant

Energy Technology Data Exchange (ETDEWEB)

Kelly, D.L.; Auflick, J.L.; Haney, L.N. [EG and G Idaho, Inc., Idaho Falls, ID (United States)

1992-04-01

Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISLOCA core damage frequency and risk. This report presents a detailed description of the application of this analysis methodology to a Westinghouse four-loop ice condenser plant. This document also includes appendices A through I which provide: System descriptions; ISLOCA event trees; human reliability analysis; thermal hydraulic analysis; core uncovery timing calculations; calculation of system rupture probability; ISLOCA consequences analysis; uncertainty analysis; and component failure analysis.

Assessment of ISLOCA risk: Methodology and application to a Westinghouse four-loop ice condenser plant

International Nuclear Information System (INIS)

Kelly, D.L.; Auflick, J.L.; Haney, L.N.

1992-04-01

Inter-system loss-of-coolant accidents (ISLOCAs) have been identified as important contributors to offsite risk for some nuclear power plants. A methodology has been developed for identifying and evaluating plant-specific hardware designs, human factors issues, and accident consequence factors relevant to the estimation of ISLOCA core damage frequency and risk. This report presents a detailed description of the application of this analysis methodology to a Westinghouse four-loop ice condenser plant. This document also includes appendices A through I which provide: System descriptions; ISLOCA event trees; human reliability analysis; thermal hydraulic analysis; core uncovery timing calculations; calculation of system rupture probability; ISLOCA consequences analysis; uncertainty analysis; and component failure analysis

[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)

A methodology for supporting decisions on the establishment of protective measures after severe nuclear accidents

International Nuclear Information System (INIS)

Papazoglou, I.A.; Kollas, J.G.

1994-06-01

The objective of this report is to demonstrate the use of a methology supporting decisions on protective measures following severe nuclear accidents. A multicriteria decision analysis approach is recommended where value tradeoffs are postponed until the very last stage of the decision process. Use of efficient frontiers is made to exclude all technically inferior solutions and present the decision maker with all nondominated solutions. A choice among these solutions implies a value trade-off among the multiple criteria. An interactive computer packge has been developed where the decision maker can choose a point on the efficient frontier in the consequence space and immediately see the alternative in the decision space resulting in the chosen consequences. The methodology is demonstrated through an application on the choice among possible protective measures in contaminated areas of the former USSR after the Chernobyl accident. Two distinct cases are considered: First a decision is to be made only on the basis of the level of soil contamination with Cs-137 and the total cost of the chosen protective policy; Next the decision is based on the geographic dimension of the contamination ant the total cost. Three alternative countermeasure actions are considered for population segments living on soil contaminated at a certain level or in a specific geographic region: (a) relocation of the population; (b) improvement of the living conditions; and, (c) no countermeasures at all. This is final deliverable of the CEC-CIS Joint Study Project 2, Task 5: Decision-Aiding-System for Establishing Intervention Levels, performed under Contracts COSU-CT91-0007 and COSU-CT92-0021 with the Commission of European Communities through CEPN

Anthropotechnological analysis of industrial accidents in Brazil.

Science.gov (United States)

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

1999-01-01

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

MELCOR DB Construction for the Severe Accident Analysis DB

International Nuclear Information System (INIS)

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

2011-01-01

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

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

Science.gov (United States)

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

2018-04-01

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

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

International Nuclear Information System (INIS)

Tong Lili; Yuan Kai; Yuan Jingtian; Cao Xuewu

2010-01-01

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

Severe accident analysis code Sampson for impact project

International Nuclear Information System (INIS)

Hiroshi, Ujita; Takashi, Ikeda; Masanori, Naitoh

2001-01-01

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

INDUSTRIAL/MILITARY ACTIVITY-INITIATED ACCIDENT SCREENING ANALYSIS

Energy Technology Data Exchange (ETDEWEB)

D.A. Kalinich

1999-09-27

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

Domino effect in chemical accidents: main features and accident sequences

OpenAIRE

Casal Fàbrega, Joaquim; Darbra Roman, Rosa Maria

2010-01-01

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

Accident and emergency management

International Nuclear Information System (INIS)

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

1988-06-01

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

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

International Nuclear Information System (INIS)

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

2006-01-01

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

Analysis of Adolescent Awareness of Radiation: Marking the First Anniversary of the Fukushima Nuclear Accident

International Nuclear Information System (INIS)

Park, Bang Ju

2012-01-01

Marking the first anniversary of the Fukushima nuclear accident, which took place on March 11th, 2011, the level of adolescent awareness and understanding of radiation was surveyed, and the results were then compared with those for adults with the same questionnaires conducted at similar times. A qualitative survey and frequency analysis were made for the design of the study methodology. Those surveyed were limited to 3rd grade middle school students, 15 years of age, who are the future generation. The questionnaire, which is a survey tool, was directly distributed to the students and 2,217 answers were analysed. The questionnaires were composed of 40 questions, and it was found that Cronbach's coefficient was high with 'self awareness of radiation' at 0.494, 'risk of radiation' at 0.843, 'benefit of radiation' at 0.748, 'radiological safety control' at 0.692, 'information sources of radiation' at 0.819, and 'impacts of Fukushima accident'. The results of the survey analysis showed that the students' knowledge of radiation was not very high with 67.4 points (69.5 points for adults) calculated on a maximum scale of 100 points (converted points). The impacts of the Fukushima nuclear accident were found to be less significant to adolescents than adults, and the rate of answer of 'so' or ' very so' in the following questions demonstrates this well. It was also shown that the impacts of the Fukushima accident to adolescents were comparatively low with 27.0% (38.9% for adults) on the question of 'attitude changed against nuclear power due to the Fukushima accident,' 65.7%(86.6% for adults) on the question of 'the damages from the Fukushima accident was immeasurably huge,' and 65.0% (86.3% for adults) on 'the Fukushima accident contributed to raising awareness on the safety of nuclear power plants'. The adolescents had a high rate of 'average' answers on most of the questions compared with adults, and it can be construed that this resulted from adolescent awareness of

Analysis of tritium mission FMEF/FAA fuel handling accidents

Energy Technology Data Exchange (ETDEWEB)

Van Keuren, J.C.

1997-11-18

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

CINETHICA - Core accident analysis code

International Nuclear Information System (INIS)

Nakata, H.

1989-10-01

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

Accident analysis for US fast burst reactors

International Nuclear Information System (INIS)

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

1994-01-01

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

Power Excursion Accident Analysis of Research Water Reactor

International Nuclear Information System (INIS)

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

2009-01-01

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

Probabilistic accident sequence recovery analysis

International Nuclear Information System (INIS)

Stutzke, Martin A.; Cooper, Susan E.

2004-01-01

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

Framatome-ANP France UO2 fuel fabrication. Criticality safety analysis in the light of the JCO accident

International Nuclear Information System (INIS)

Doucet, M.; Zheng, S.; Mouton, J.; Porte, R.

2003-01-01

In France the 1999' Tokai Mura criticality accident in Japan had a big impact on the nuclear fuel manufacturing facility community. Moreover this accident led to a large public discussion about all the nuclear facilities. The French Safety Authorities made strong requirements to the industrials to revisit completely their safety analysis files mainly those concerning nuclear fuels treatments. The FRAMATOME-ANP production of its French low enriched (5 w/o) UO2 fuel fabrication plant (FBFC/Romans) exceeds 1000 metric tons a year. Special attention was given to the emergency evacuation plan that should be followed in case of a criticality accident. If a criticality accident happens, site internal and external radioprotection requirements need to have an emergency evacuation plan showing the different routes where the absorbed doses will be as low as possible for people. The French Safety Authorities require also an update of the old based neutron source term accounting for state of the art methodology. UO2 blenders units contain a large amount of dry powder strictly controlled by moderation; a hypothetical water leakage inside one of these apparatus is simulated by increasing the water content of the powder. The resulted reactivity insertion is performed by several static calculations. The French IRSN/CEA CRISTAL codes are used to perform these static calculations. The kinetic criticality code POWDER simulates the power excursion versus time and determines the consequent total energy source term. MNCP4B performs the source term propagation (including neutrons and gamma) used to determine the isodose curves needed to define the emergency evacuation plant. This paper deals with the approach FRAMATOME-ANP has taken to assess Safety Authorities demands using the more up to date calculation tools and methodology. (author)

Human factors review for Severe Accident Sequence Analysis (SASA)

International Nuclear Information System (INIS)

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

1984-01-01

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

CANDU severe accident analysis

International Nuclear Information System (INIS)

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

2007-01-01

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

A System Supporting the Analysis of Motorway Traffic Accidents

Directory of Open Access Journals (Sweden)

Davide Anghinolfi

2015-12-01

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

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

Science.gov (United States)

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

2017-10-01

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

Risk Analysis of Fukushima Accident using MACCS2

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2012-09-30

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

National and regional analysis of road accidents in Spain.

Science.gov (United States)

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

2013-01-01

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

Analysis of Child-related Road Traffic Accidents in Vietnam

Science.gov (United States)

Vu, Anh Tuan; Nguyen, Dinh Vinh Man

2018-04-01

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

An Analysis of Construction Accident Factors Based on Bayesian Network

OpenAIRE

Yunsheng Zhao; Jinyong Pei

2013-01-01

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

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

Directory of Open Access Journals (Sweden)

Arleiny

2018-01-01

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

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

Science.gov (United States)

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

2018-01-01

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

Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

International Nuclear Information System (INIS)

Chang, Y.H.; Mosleh, A.; Dang, V.N.

2003-01-01

The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

Development Of Dynamic Probabilistic Safety Assessment: The Accident Dynamic Simulator (ADS) Tool

Energy Technology Data Exchange (ETDEWEB)

Chang, Y.H.; Mosleh, A.; Dang, V.N

2003-03-01

The development of a dynamic methodology for Probabilistic Safety Assessment (PSA) addresses the complex interactions between the behaviour of technical systems and personnel response in the evolution of accident scenarios. This paper introduces the discrete dynamic event tree, a framework for dynamic PSA, and its implementation in the Accident Dynamic Simulator (ADS) tool. Dynamic event tree tools generate and quantify accident scenarios through coupled simulation models of the plant physical processes, its automatic systems, the equipment reliability, and the human response. The current research on the framework, the ADS tool, and on Human Reliability Analysis issues within dynamic PSA, is discussed. (author)

Studies of potential severe accidents in Finnish nuclear power plants. Quarterly report 3. quarter 1987

International Nuclear Information System (INIS)

Aro, Ilari.

1989-07-01

This thesis is based on six publications dealing with severe accident studies in Finnish nuclear power plants. Main emphasis has been put on general technical bases and methodologies applied in severe accident evaluation in Finland. As an example of the use of the analysis and evaluation methods, the analysis of one representative accident sequence, t otal loss of AC power , has been presented for both Finnish power plant types. This accident sequence is required to be analyzed in the Finnish safety guide YVL 2.2 which deals with transient and accident analyses as a basis of technical solutions at nuclear powr plants. Two different analysis methods, MAAP 3.0 and MARCH 3/STCP have been used for receiving as complete a picture as possible of the flow of events and for verifying the models to some extent. Besides the use of the two different models, the method of sensitivity analysis has been used for evaluating the effects of some important technical parameters on the accident flow. Finally, conclusions of the applicability of the two methods for analyzing severe accident sequences in Finnish plants have been discussed

Reactor Core Coolability Analysis during Hypothesized Severe Accidents of OPR1000

International Nuclear Information System (INIS)

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

2014-01-01

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

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

Science.gov (United States)

Kumar, R; Ghosh, A K

2014-01-01

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

Accident rate analysis for well drilling at Kuban' morneftegazprom association

Energy Technology Data Exchange (ETDEWEB)

Sukhanov, V B; Kezchikov, A V

1981-01-01

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

A critical assessment of energy accident studies

International Nuclear Information System (INIS)

Felder, Frank A.

2009-01-01

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases.

A critical assessment of energy accident studies

Energy Technology Data Exchange (ETDEWEB)

Felder, Frank A. [Edward J. Bloustein School of Planning and Public Policy, Rutgers, The State University of New Jersey, 33 Livingston Avenue, New Brunswick, NJ 08901 (United States)

2009-12-15

A comparison of two studies conducted ten years apart on energy accidents provides important insights into methodological issues and policy implications. Recommendations for further improvements in energy accident studies are developed including accounting for differences between average and incremental accident damages, testing for appropriate levels of aggregation of accidents, making references and databases publicly available, more precisely defining and reporting different types of economic damages, accounting for involuntary and voluntary risks, reporting normalized damages, raising broader public policy and planning implications and updating existing accident databases. (author)

IMPROVING METHODOLOGY OF RISK IDENTIFICATION OF OCCUPATIONAL DANGEROUS

Directory of Open Access Journals (Sweden)

A. P. BOCHKOVSKYI

2018-04-01

Full Text Available In the paper, according to the analysis of statistical data, correlation between the amount of occupational injuries and occupationaldiseases in Ukraine within last 5 years is defined. Also, using methodology of the International Labor Organizationcorrelcation between the amount of accident fatalities and general number of accidents in Ukraine and EU countries (Austria, GreatBritain, Germany, Denmark, Norway, Poland, Hungry, Finland, France is defined. It is shown that in spite of the positive dynamicsof decreasing amount of occupational injuries, the number of occupational diseases in Ukraine always increases. The comparativeanalysis of the ratio of the number of accident fatalities to the total number of registered accidents showed that, on average, Ukraineexceeds the EU countries by this indicator by 100 times.It is noted, that such negative indicators (in particular, increasing amount of occupational diseases, may occure because ofimperfect methodology for identifying the risks of professional dangerous.Also, it is ascertained that basing on the existed methodology, the identefication process of occupational dangerous isquite subjective, which reduces objectivity of conducting quantitative assessment. In order to eliminate defined drawnbacks it is firsttime proposed to use corresponding integral criterion to conduct the process of quantitative risk assessmentTo solve this problem authors formulate and propose an algorithm of improving methodology of a process of analysing dangerousand harmful production effects (DHPE which are the mainest reasons of occupational dangerous.The proposed algorithm includes implementation of four following successive steps: DHPE identification, indication of theirmaximum allowed threshold of concentrations (levels, identification of the sources of identified DHPE, esimation of consequencesof manifestation.The improved proposed methodology allows indentify risks of occurrence occupational dangerous in systems

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

International Nuclear Information System (INIS)

Park, Soo Young; Ahn, Kwang Il

2012-01-01

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

Cognitive systems engineering analysis of the JCO criticality accident

International Nuclear Information System (INIS)

Tanabe, Fumiya; Yamaguchi, Yukichi

2000-01-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

2017-10-01

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

Fuel temperature analysis method for channel-blockage accident in HTTR

International Nuclear Information System (INIS)

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

1994-01-01

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

Regional Shelter Analysis Methodology

Energy Technology Data Exchange (ETDEWEB)

Dillon, Michael B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Dennison, Deborah [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Kane, Jave [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Walker, Hoyt [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Miller, Paul [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

2015-08-01

The fallout from a nuclear explosion has the potential to injure or kill 100,000 or more people through exposure to external gamma (fallout) radiation. Existing buildings can reduce radiation exposure by placing material between fallout particles and exposed people. Lawrence Livermore National Laboratory was tasked with developing an operationally feasible methodology that could improve fallout casualty estimates. The methodology, called a Regional Shelter Analysis, combines the fallout protection that existing buildings provide civilian populations with the distribution of people in various locations. The Regional Shelter Analysis method allows the consideration of (a) multiple building types and locations within buildings, (b) country specific estimates, (c) population posture (e.g., unwarned vs. minimally warned), and (d) the time of day (e.g., night vs. day). The protection estimates can be combined with fallout predictions (or measurements) to (a) provide a more accurate assessment of exposure and injury and (b) evaluate the effectiveness of various casualty mitigation strategies. This report describes the Regional Shelter Analysis methodology, highlights key operational aspects (including demonstrating that the methodology is compatible with current tools), illustrates how to implement the methodology, and provides suggestions for future work.

Multi-objective evolutionary emergency response optimization for major accidents

International Nuclear Information System (INIS)

Georgiadou, Paraskevi S.; Papazoglou, Ioannis A.; Kiranoudis, Chris T.; Markatos, Nikolaos C.

2010-01-01

Emergency response planning in case of a major accident (hazardous material event, nuclear accident) is very important for the protection of the public and workers' safety and health. In this context, several protective actions can be performed, such as, evacuation of an area; protection of the population in buildings; and use of personal protective equipment. The best solution is not unique when multiple criteria are taken into consideration (e.g. health consequences, social disruption, economic cost). This paper presents a methodology for multi-objective optimization of emergency response planning in case of a major accident. The emergency policy with regards to protective actions to be implemented is optimized. An evolutionary algorithm has been used as the optimization tool. Case studies demonstrating the methodology and its application in emergency response decision-making in case of accidents related to hazardous materials installations are presented. However, the methodology with appropriate modification is suitable for supporting decisions in assessing emergency response procedures in other cases (nuclear accidents, transportation of hazardous materials) or for land-use planning issues.

Formulating accident occurrence as a survival process.

Science.gov (United States)

Chang, H L; Jovanis, P P

1990-10-01

A conceptual framework for accident occurrence is developed based on the principle of the driver as an information processor. The framework underlies the development of a modeling approach that is consistent with the definition of exposure to risk as a repeated trial. Survival theory is proposed as a statistical technique that is consistent with the conceptual structure and allows the exploration of a wide range of factors that contribute to highway operating risk. This survival model of accident occurrence is developed at a disaggregate level, allowing safety researchers to broaden the scope of studies which may be limited by the use of traditional aggregate approaches. An application of the approach to motor carrier safety is discussed as are potential applications to a variety of transportation industries. Lastly, a typology of highway safety research methodologies is developed to compare the properties of four safety methodologies: laboratory experiments, on-the-road studies, multidisciplinary accident investigations, and correlational studies. The survival theory formulation has a mathematical structure that is compatible with each safety methodology, so it may facilitate the integration of findings across methodologies.

Evaluation of severe accident risks: Methodology for the containment, source term, consequence, and risk integration analyses. Volume 1, Revision 1

International Nuclear Information System (INIS)

Gorham, E.D.; Breeding, R.J.; Brown, T.D.; Harper, F.T.; Helton, J.C.; Murfin, W.B.; Hora, S.C.

1993-12-01

NUREG-1150 examines the risk to the public from five nuclear power plants. The NUREG-1150 plant studies are Level III probabilistic risk assessments (PRAs) and, as such, they consist of four analysis components: accident frequency analysis, accident progression analysis, source term analysis, and consequence analysis. This volume summarizes the methods utilized in performing the last three components and the assembly of these analyses into an overall risk assessment. The NUREG-1150 analysis approach is based on the following ideas: (1) general and relatively fast-running models for the individual analysis components, (2) well-defined interfaces between the individual analysis components, (3) use of Monte Carlo techniques together with an efficient sampling procedure to propagate uncertainties, (4) use of expert panels to develop distributions for important phenomenological issues, and (5) automation of the overall analysis. Many features of the new analysis procedures were adopted to facilitate a comprehensive treatment of uncertainty in the complete risk analysis. Uncertainties in the accident frequency, accident progression and source term analyses were included in the overall uncertainty assessment. The uncertainties in the consequence analysis were not included in this assessment. A large effort was devoted to the development of procedures for obtaining expert opinion and the execution of these procedures to quantify parameters and phenomena for which there is large uncertainty and divergent opinions in the reactor safety community

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

Directory of Open Access Journals (Sweden)

Mohamadfam

2015-01-01

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

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

Science.gov (United States)

Underwood, Peter; Waterson, Patrick

2013-06-01

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

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

International Nuclear Information System (INIS)

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

1985-01-01

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

Work accidents during cable yarding operations in Central Europe 2006 – 2014

International Nuclear Information System (INIS)

Allman, M.; Jankovský, M.; Allmanová, Z.; Ferenčík, M.; Messingerová, V.; Vlčková, M.; Stanimir Stoilov

2017-01-01

Aim of study: This study is focused on detailed analysis of accidents in yarding during the years 2006–2014. There is still not enough information about such accidents in Central Europe in the literature available. Area of study: We collected the data on occupational accidents recorded in timber yarding from the databases of the Slovak state forest enterprise. Material and Methods: The data on occupational accidents were recorded according to actual European Regulation, the form of the record meets the requirements of the ESAW (European Statistics on Accidents at Work) methodology. To analyze the data, we used the multiple regression and correlation analysis, contingency tables, and a χ2 –test. Main results: Almost half of the accidents were the foot injuries and the most frequent type of injury was fracture of a bone. The most hazardous operation was yarding. Most of the accidents occurred between 1301-1400 h (22 %). The most frequent agent causing accidents were Particles, dust, splinters, fragments, etc. (14.05 by ESAW). Research highlights: This study informs about the most important risk factors in timber yarding, the most hazardous parts of shift, as well as the days when the most accidents occur during the week, and as such contributes to better understanding of how the accidents happen in timber yarding. The information can be subsequently used in knowledge-based improvement of safety trainings in forest enterprises.

Work accidents during cable yarding operations in Central Europe 2006 – 2014

Energy Technology Data Exchange (ETDEWEB)

Allman, M.; Jankovský, M.; Allmanová, Z.; Ferenčík, M.; Messingerová, V.; Vlčková, M.; Stanimir Stoilov

2017-11-01

Aim of study: This study is focused on detailed analysis of accidents in yarding during the years 2006–2014. There is still not enough information about such accidents in Central Europe in the literature available. Area of study: We collected the data on occupational accidents recorded in timber yarding from the databases of the Slovak state forest enterprise. Material and Methods: The data on occupational accidents were recorded according to actual European Regulation, the form of the record meets the requirements of the ESAW (European Statistics on Accidents at Work) methodology. To analyze the data, we used the multiple regression and correlation analysis, contingency tables, and a χ2 –test. Main results: Almost half of the accidents were the foot injuries and the most frequent type of injury was fracture of a bone. The most hazardous operation was yarding. Most of the accidents occurred between 1301-1400 h (22 %). The most frequent agent causing accidents were Particles, dust, splinters, fragments, etc. (14.05 by ESAW). Research highlights: This study informs about the most important risk factors in timber yarding, the most hazardous parts of shift, as well as the days when the most accidents occur during the week, and as such contributes to better understanding of how the accidents happen in timber yarding. The information can be subsequently used in knowledge-based improvement of safety trainings in forest enterprises.

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

Science.gov (United States)

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

2014-12-01

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

Development of Severe Accident Containment Analysis Package

Energy Technology Data Exchange (ETDEWEB)

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

2016-10-15

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

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

NARCIS (Netherlands)

Oppe, S.

1982-01-01

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

Hanford Waste Tank Bump Accident and Consequence Analysis

International Nuclear Information System (INIS)

BRATZEL, D.R.

2000-01-01

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

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

International Nuclear Information System (INIS)

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

2002-01-01

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

Applicability of simplified human reliability analysis methods for severe accidents

Energy Technology Data Exchange (ETDEWEB)

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

2016-03-15

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

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

DEFF Research Database (Denmark)

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

2010-01-01

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

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

International Nuclear Information System (INIS)

Fontes, S.W.F.

1989-01-01

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

Current status of low power/shutdown PSA and accident sequence analysis for loss of RHR during mid-loop operation

International Nuclear Information System (INIS)

Park, Chang Kyu; Choi, Young; Kim, Tae Woon; Jin, Young Ho

1994-07-01

Probabilistic safety assessment (PSA) has been applied to only full-power operation of nuclear power plant (NPP), but some events which were recently occurred could reach severe plant damage state. Thus, various countries around the world have focused their interests on the evaluation for low power/shutdown (LP/S) operation. This report covers the main stream of LP/S PSA methodology, current status of LP/S PSA practices and results, and accident sequence analysis for loss of RHR during mid-loop operation. Therefore this report would be helpful for us to practice LP/S PSA for YGN 5,6 NPP which will be built in the near future. Also the results of accident sequence analysis show that operator's mis-diagnosis and failure of recovery action would initiate core damage during LP/S operation. In summary, overall environmental improvements (equipments, procedures, Tech Spec, etc, ...) and operating support system will be very useful to reduce risk during LP/S operation. (Author) 5 figs., 9 tabs

Qualification and application of nuclear reactor accident analysis code with the capability of internal assessment of uncertainty

International Nuclear Information System (INIS)

Borges, Ronaldo Celem

2001-10-01

This thesis presents an independent qualification of the CIAU code ('Code with the capability of - Internal Assessment of Uncertainty') which is part of the internal uncertainty evaluation process with a thermal hydraulic system code on a realistic basis. This is done by combining the uncertainty methodology UMAE ('Uncertainty Methodology based on Accuracy Extrapolation') with the RELAP5/Mod3.2 code. This allows associating uncertainty band estimates with the results obtained by the realistic calculation of the code, meeting licensing requirements of safety analysis. The independent qualification is supported by simulations with RELAP5/Mod3.2 related to accident condition tests of LOBI experimental facility and to an event which has occurred in Angra 1 nuclear power plant, by comparison with measured results and by establishing uncertainty bands on safety parameter calculated time trends. These bands have indeed enveloped the measured trends. Results from this independent qualification of CIAU have allowed to ascertain the adequate application of a systematic realistic code procedure to analyse accidents with uncertainties incorporated in the results, although there is an evident need of extending the uncertainty data base. It has been verified that use of the code with this internal assessment of uncertainty is feasible in the design and license stages of a NPP. (author)

Human factors analysis of incident/accident report

International Nuclear Information System (INIS)

Kuroda, Isao

1992-01-01

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

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

Cost-effectiveness analysis of countermeasures using accident consequence assessment models

International Nuclear Information System (INIS)

Alonso, A.; Gallego, E.

1987-01-01

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

Accident analysis of Fukushima Daiichi Nuclear Power Station unit 1

International Nuclear Information System (INIS)

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

2015-01-01

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

A trend analysis of human error events for proactive prevention of accidents. Methodology development and effective utilization

International Nuclear Information System (INIS)

Hirotsu, Yuko; Ebisu, Mitsuhiro; Aikawa, Takeshi; Matsubara, Katsuyuki

2006-01-01

This paper described methods for analyzing human error events that has been accumulated in the individual plant and for utilizing the result to prevent accidents proactively. Firstly, a categorization framework of trigger action and causal factors of human error events were reexamined, and the procedure to analyze human error events was reviewed based on the framework. Secondly, a method for identifying the common characteristics of trigger action data and of causal factor data accumulated by analyzing human error events was clarified. In addition, to utilize the results of trend analysis effectively, methods to develop teaching material for safety education, to develop the checkpoints for the error prevention and to introduce an error management process for strategic error prevention were proposed. (author)

A Methodology for Safety Culture Impact Assessment

Energy Technology Data Exchange (ETDEWEB)

Han, Kiyoon; Jae, Moosung [Hanyang Univ., Seoul (Korea, Republic of)

2014-05-15

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively.

A Methodology for Safety Culture Impact Assessment

International Nuclear Information System (INIS)

Han, Kiyoon; Jae, Moosung

2014-01-01

The purpose of this study is to develop methodology for assessing safety culture impact on nuclear power plants. A new methodology for assessing safety culture impact index has been developed and applied for the reference nuclear power plants. The developed SCII model might contribute to comparing the level of safety culture among nuclear power plants as well as to improving the safety of nuclear power plants. Safety culture is defined to be fundamental attitudes and behaviors of the plant staff which demonstrate that nuclear safety is the most important consideration in all activities conducted in nuclear power operation. Through several accidents of nuclear power plant including the Fukusima Daiichi in 2011 and Chernovyl accidents in 1986, the safety of nuclear power plant is emerging into a matter of interest. From the accident review report, it can be easily found out that safety culture is important and one of dominant contributors to accidents. However, the impact methodology for assessing safety culture has not been established analytically yet. It is difficult to develop the methodology for assessing safety culture impact quantitatively

Accident analysis for transuranic waste management alternatives in the U.S. Department of Energy waste management program

International Nuclear Information System (INIS)

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

1995-01-01

Preliminary accident analyses and radiological source term evaluations have been conducted for transuranic waste (TRUW) as part of the US Department of Energy (DOE) effort to manage storage, treatment, and disposal of radioactive wastes at its various sites. The approach to assessing radiological releases from facility accidents was developed in support of the Office of Environmental Management Programmatic Environmental Impact Statement (EM PEIS). The methodology developed in this work is in accordance with the latest DOE guidelines, which consider the spectrum of possible accident scenarios in the implementation of various actions evaluated in an EIS. The radiological releases from potential risk-dominant accidents in storage and treatment facilities considered in the EM PEIS TRUW alternatives are described in this paper. The results show that significant releases can be predicted for only the most severe and extremely improbable accidents sequences

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

Energy Technology Data Exchange (ETDEWEB)

Hanson, Duane J. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Arcieri, William C. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States)); Ward, Leonard W. (Idaho National Engineering Laboratory, Idaho Falls, ID 83415 (United States))

1994-07-01

A five-step methodology was 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 that 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 severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

Assessing information needs and instrument availability for a pressurized water reactor during severe accidents

International Nuclear Information System (INIS)

Hanson, Duane J.; Arcieri, William C.; Ward, Leonard W.

1994-01-01

A five-step methodology was 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 that 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 severe accident conditions, to evaluate the availability of the instrumentation to supply needed plant information. This methodology was applied to a pressurized water reactor with a large dry containment and the results are presented. A companion article describes application of the methodology to a boiling water reactor with a Mark I containment. ((orig.))

Human factors review for nuclear power plant severe accident sequence analysis

International Nuclear Information System (INIS)

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

1985-01-01

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

The ARIPAR project: analysis of the major accident risks connected with industrial and transportation activities in the Ravenna area

International Nuclear Information System (INIS)

Egidi, Demetrio; Foraboschi, Franco P.; Spadoni, Gigliola; Amendola, Aniello

1995-01-01

The paper describes the ARIPAR project aimed at the assessment of the major accident risks connected with storage, process and transportation of dangerous substances in the densely populated Ravenna area in Italy, which includes a large complex of chemical and petrochemical plants and minor industries, essentially distributed around an important commercial port. Large quantities of dangerous goods are involved in various transportation forms connected with the industrial and commercial activity of the port. The project started by making a complete inventory of fixed installations and transportation activities capable of provoking major fire, explosion and toxic release events; then relevant accident scenarios were developed for the single hazard sources; probabilities were assigned to the events and consequences were evaluated; finally iso-risk contours and F-N diagrams were evaluated both for the single sources and for the overall area. This required the development of a particular methodology for analysis of area risk and of associated software packages which allowed examination of the relative importance of the different activities and typologies of materials involved. The methodological approach and the results have proved to be very useful for the priority-ranking of risk mitigating interventions and physical planning in a complex area

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

International Nuclear Information System (INIS)

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

1996-01-01

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

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

OpenAIRE

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

2013-01-01

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

Quantification of severe accident source terms of a Westinghouse 3-loop plant

International Nuclear Information System (INIS)

Lee Min; Ko, Y.-C.

2008-01-01

Integrated severe accident analysis codes are used to quantify the source terms of the representative sequences identified in PSA study. The characteristics of these source terms depend on the detail design of the plant and the accident scenario. A historical perspective of radioactive source term is provided. The grouping of radionuclides in different source terms or source term quantification tools based on TID-14844, NUREG-1465, and WASH-1400 is compared. The radionuclides release phenomena and models adopted in the integrated severe accident analysis codes of STCP and MAAP4 are described. In the present study, the severe accident source terms for risk quantification of Maanshan Nuclear Power Plant of Taiwan Power Company are quantified using MAAP 4.0.4 code. A methodology is developed to quantify the source terms of each source term category (STC) identified in the Level II PSA analysis of the plant. The characteristics of source terms obtained are compared with other source terms. The plant analyzed employs a Westinghouse designed 3-loop pressurized water reactor (PWR) with large dry containment

A review of the current thermal-hydraulic modeling of the Jules Horowitz Reactor: A loss of flow accident analysis

International Nuclear Information System (INIS)

Pegonen, R.; Bourdon, S.; Gonnier, C.; Anglart, H.

2014-01-01

Highlights: • CEA methodology for thermal-hydraulic calculations in the JHR reactor is described. • Thermal-hydraulics of the JHR is analyzed during LOFA using CATHARE and FLICA4. • Safety criteria, important modeling parameters and correlations are presented. • Possible improvements of the current methodology are discussed and proposed. - Abstract: The newest European high performance material testing reactor, the Jules Horowitz Reactor, will support existing and future nuclear reactor designs. The reactor is under construction at CEA Cadarache research center in France and is expected to start operation at the end of this decade. R and D and analytical works have already been performed to set-up the methodology for thermal-hydraulic calculations of the reactor. This paper presents the off-line coupled thermal-hydraulic modeling of the reactor using the CATHARE system code and the FLICA4 core analysis code. The main objective of the present work is to analyze the thermal-hydraulic calculations of the reactor during the loss of flow accident using CEA methodology. Possible improvements of the current methodology are shortly discussed and suggested

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

OpenAIRE

FABBRI LUCIANO; BINDA MASSIMO; BRUINEN DE BRUIN YURI

2017-01-01

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

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

International Nuclear Information System (INIS)

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

1998-01-01

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

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)

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

Science.gov (United States)

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

2014-01-01

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