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. 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. Compes, P.C. 1987-01-01 The example of the Chernobyl accident and the statistics of the occurrence of accidents make clear the threat to humanity, if one cannot guarantee successful accident prevention in the use and distribution of the projects aimed at.
The science of safety, as it is known in the Wuppertal model, makes its contribution to this vital task for the human community. It makes it necessary to create the essential dates and concepts, the methods, principles and techniques based on them and the associated instrumentation. (DG) de. 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).
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. 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. 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).
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. 1975-10-01 Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks. Galyean, W.J.; Kelly, D.L.; Schroeder, J.A.; Auflick, L.J.; Blackman, H.S.; Gertman, D.I.; Hanley, L.N. 1994-01-01 The United States Nuclear Regulatory Commission is sponsoring a research program to develop an improved understanding of the human factors, hardware and accident consequence issues that dominate the risk from an intersystem loss-of-coolant accident (ISLOCA) at a nuclear power plant. To accomplish the goals of this program, a mehtodology has been developed for estimating ISLOCA core damage frequency and risk.
The steps in this methodology are briefly described, along with the results obtained from an application of the methodology at three pressurized water reactors. Also included are the results of a screening study of boiling water reactors. ((orig.)). 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. 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. Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E. 1989-01-01 The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents.
Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel. Ward, L.W.; Hanson, D.J.; Nelson, W.R. (Idaho National Engineering Laboratory, Idaho Falls (USA)); Solberg, D.E.
(Nuclear Regulatory Commission, Washington, DC (USA)) 1989-11-01 The U.S. Nuclear Regulatory Commission is conducting an accident management research program that emphasizes the use of severe accident research to enhance the ability of plant operating personnel to effectively manage severe accidents. Hence, it is necessary to ensure that the plant instrumentation and information systems adequately provide this information to the operating staff during accident conditions.
A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed. The methodology identifies (a) the information needs of the plant personnel during a wide range of accident conditions, (b) the existing plant measurements capable of supplying these information needs and minor additions to instrument and display systems that would enhance management capabilities, (c) measurement capabilities and limitations during severe accident conditions, and (d) areas in which the information systems could mislead plant personnel. 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).
Ward, L.W.; Hanson, D.J.; Nelson, W.R.; Solberg, D.E. 1989-01-01 The U.S. Nuclear Regulatory Commission (NRC) is conducting an Accident Management Research Program that emphasizes the application of severe accident research results to enhance the capability of plant operating personnel to effectively manage severe accidents. A methodology to identify and assess the information needs of the operating staff of a nuclear power plant during a severe accident has been developed as part of the research program designed to resolve this issue.
The methodology identifies the information needs of the plant personnel during a wide range of accident conditions, the existing plant measurements capable of supplying these information needs and what, if any minor additions to instrument and display systems would enhance the capability to manage accidents, known limitations on the capability of these measurements to function properly under the conditions that will be present during a wide range of severe accidents, and areas in which the information systems could mislead plant personnel. This paper presents an application of this methodology to a severe accident sequence to demonstrate its use in identifying the information which is available for management of the event. The methodology has been applied to a severe accident sequence in a Pressurized Water Reactor with a large dry containment. An examination of the capability of the existing measurements was then performed to determine whether the information needs can be supplied.
Silva, F.C.A. 1990-01-01 A critical review of 34 published severe radiological accidents in industrial gamma radiography, that happened in 15 countries, from 1960 to 1988, was performed.
The most frequent causes, consequences and dose estimation methods were analysed, aiming to stablish better procedures of radiation safety and accidents analysis. The objective of this work is to elaborate a radiological accidents analysis methodology in industrial gamma radiography. The suggested methodology will enable professionals to determine the true causes of the event and to estimate the dose with a good certainty. The technical analytical tree, recommended by International Atomic Energy Agency to perform radiation protection and nuclear safety programs, was adopted in the elaboration of the suggested methodology. The viability of the use of the Electron Gamma Shower 4 Computer Code System to calculate the absorbed dose in radiological accidents in industrial gamma radiography, mainly at sup(192)Ir radioactive source handling situations was also studied.
(author). 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.
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). Bennett, A; Parks, S 2006-04-01 To quantify microbial aerosols generated by a series of laboratory accidents and to use these data in risk assessment. A series of laboratory accident scenarios have been devised and the microbial aerosol generated by them has been measured using a range of microbial air samplers.
The accident scenarios generating the highest aerosol concentrations were, dropping a fungal plate, dropping a large bottle, centrifuge rotor leaks and a blocked syringe filter. Many of these accidents generated low particle size aerosols, which would be inhaled into the lungs of any exposed laboratory staff. Spray factors (SFs) have been calculated using the results of these experiments as an indicator of the potential for accidents to generate microbial aerosols.
Model risk assessments have been described using the SF data. Quantitative risk assessment of laboratory accidents can provide data that can aid the design of containment laboratories and the response to laboratory accidents.
A methodology has been described and supporting data provided to allow microbiological safety officers to carry out quantitative risk assessment of laboratory accidents. 1975-10-01 Information is presented concerning the objectives and organization of the reactor safety study; the basic concepts of risk; the nature of nuclear power plant accidents; risk assessment methodology; reactor accident risk; and comparison of nuclear risks to other societal risks. 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. Keizer, J.A.; Halman, J.I.M.; Song, X.M. 2002-01-01 No risk, no reward. Companies must take risks to launch new products speedily and successfully. The ability to diagnose and manage risks is increasingly considered of vital importance in high- risk innovation.
This article presents the Risk Diagnosing Methodology (RDM), which aims to identify and. Keizer, Jimme A.; Halman, Johannes I.M.; Song, Michael 2002-01-01 No risk, no reward.
Companies must take risks to launch new products speedily and successfully. The ability to diagnose and manage risks is increasingly considered of vital importance in high- risk innovation. This article presents the Risk Diagnosing Methodology (RDM), which aims to identify and. 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. Bott, T.F.; Mac Farlane, D.R.; Stack, D.W.; Kindinger, J. 1992-01-01 A methodology is presented for applying Probabilistic Safety Assessment techniques to quantification of the health risks posed by the high-level waste (HLW) underground tanks at the Department of Energy's Hanford reservation. This methodology includes hazard screening development of a list of potential accident initiators, systems fault trees development and quantification, definition of source terms for various release categories, and estimation of health consequences from the releases.
Both airborne and liquid pathway releases to the environment, arising from aerosol and spill/leak releases from the tanks, are included in the release categories. The proposed methodology is intended to be applied to a representative subset of the total of 177 tanks, thereby providing a baseline risk profile for the HLW tank farm that can be used for setting clean-up/remediation priorities. Some preliminary results are presented for Tank 101-SY.
Wesley, D.A. 1991-01-01 A loss of coolant accident resulting from the overpressurization by reactor coolant fluid of a system designed for low-pressure, low-temperature service has been identified as a potential contributor to nuclear power plant risk. In this paper, the methodology developed to assess the probability of failure as a function of internal pressure is presented, and sample results developed for the controlling failure modes and locations of four fluid systems at the Davis-Besse Plant are shown.
Included in this evaluation are the tanks, heat exchangers, filters, pumps, valves, and flanged connections for each system. The variability in the probability of failure is included, and the estimated leak rates or leak areas are given for the controlling modes of failure. For this evaluation, all failures are based on quasistatic pressures since the probability of dynamic effects resulting from such causes as water hammer have been initially judged to be negligible for the Davis-Besse plant ISLOCA.
Lindell, B 1986-01-01 A review is given of the basic of radiation protection, including nomenclature and units and principles for protection at accidents. The consequences of the Chernobyl accident in the Soviet Union and in Sweden is described, and the recommendations and protection measures applied in Sweden are presented.
In particular, the radiation levels and restrictions concerning food are discussed. 1993-03-01 This methodology has been developed to prepare human health and environmental evaluations of risk as part of the Comprehensive Environmental Response, Compensation, and Liability Act remedial investigations (RIs) and the Resource Conservation and Recovery Act facility investigations (FIs) performed at the Hanford Site pursuant to the Hanford Federal Facility Agreement and Consent Order referred to as the Tri-Party Agreement. Development of the methodology has been undertaken so that Hanford Site risk assessments are consistent with current regulations and guidance, while providing direction on flexible, ambiguous, or undefined aspects of the guidance. The methodology identifies Site-specific risk assessment considerations and integrates them with approaches for evaluating human and environmental risk that can be factored into the risk assessment program supporting the Hanford Site cleanup mission. Consequently, the methodology will enhance the preparation and review of individual risk assessments at the Hanford Site. 1995-05-01 This methodology has been developed to prepare human health and ecological evaluations of risk as part of the Comprehensive Environmental Response, Compensation, and Liability Act of 1980 (CERCLA) remedial investigations (RI) and the Resource conservation and Recovery Act of 1976 (RCRA) facility investigations (FI) performed at the Hanford Site pursuant to the hanford Federal Facility Agreement and Consent Order (Ecology et al.
1994), referred to as the Tri-Party Agreement. Development of the methodology has been undertaken so that Hanford Site risk assessments are consistent with current regulations and guidance, while providing direction on flexible, ambiguous, or undefined aspects of the guidance. The methodology identifies site-specific risk assessment considerations and integrates them with approaches for evaluating human and ecological risk that can be factored into the risk assessment program supporting the Hanford Site cleanup mission. Consequently, the methodology will enhance the preparation and review of individual risk assessments at the Hanford Site.
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. Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S., E-mail: [email protected], E-mail: [email protected], E-mail: [email protected] Instituto de Pesquisas Energeticas e Nucleares (IPEN/CNEN-SP), Sao Paulo, SP (Brazil) 2015-07-01 The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA).
This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents. One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident. The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA.
The boundary conditions for the simulation are obtained from RELAP5 code. (author). Silva, Dayane F.; Sabundjian, Gaiane; Lima, Ana Cecilia S. 2015-01-01 The aim of this work is to present the methodology of calculation to Angra 2 reactor containment during accidents of the type Loss of Coolant Accident (LOCA). This study will be possible to ensure the safety of the population of the surroundings upon the occurrence of accidents.
One of the programs used to analyze containment of a nuclear plant is the CONTAIN. This computer code is an analysis tool used for predicting the physical conditions and distributions of radionuclides inside a containment building following the release of material from the primary system in a light-water reactor during an accident.
The containment of the type PWR plant is a concrete building covered internally by metallic material and has limits of design pressure. The methodology of containment analysis must estimate the limits of pressure during a LOCA. The boundary conditions for the simulation are obtained from RELAP5 code.
(author). Luxat, J.C., E-mail: [email protected] McMaster Univ., Dept. Of Engineering Physics, Hamilton, Ontario (Canada) 2013-07-01 Major accidents and natural disasters with severe consequences have occurred in all sectors of industrial activity with relatively high frequency.
The severe consequences of concern involve either significant loss of life or major economic loss, or both loss of life and economic loss. Such events have in the last two years often been referred to as 'Black Swan' events following publication of a best-selling book. The events demonstrate limits to the application of Probabilistic Risk Assessment (PRA) that arise from the underlying unquantifiable uncertainty associated with the estimation of the frequency of occurrence of such events. An approach is proposed in this paper that, consistent with the concept of defense in depth employed by the nuclear industry, augments probabilistic risk assessment with a methodology based upon 'threat - risk assessment'. This approach shifts these very low frequency, high uncertainty, and high consequence 'Black Swan' events out of the probabilistic risk assessment domain and into a deterministic emergency response assessment domain. (author). Luxat, J.C.
2013-01-01 Major accidents and natural disasters with severe consequences have occurred in all sectors of industrial activity with relatively high frequency. The severe consequences of concern involve either significant loss of life or major economic loss, or both loss of life and economic loss. Such events have in the last two years often been referred to as 'Black Swan' events following publication of a best-selling book. The events demonstrate limits to the application of Probabilistic Risk Assessment (PRA) that arise from the underlying unquantifiable uncertainty associated with the estimation of the frequency of occurrence of such events. An approach is proposed in this paper that, consistent with the concept of defense in depth employed by the nuclear industry, augments probabilistic risk assessment with a methodology based upon 'threat - risk assessment'. This approach shifts these very low frequency, high uncertainty, and high consequence 'Black Swan' events out of the probabilistic risk assessment domain and into a deterministic emergency response assessment domain.
(author). Saša T. Bakrač 2012-04-01 Full Text Available Successful protection of environment is mostly based on high-quality assessment of potential and present risks. Environmental risk management is a complex process which includes: identification, assessment and control of risk, namely taking measures in order to minimize the risk to an acceptable level. Environmental risk management methodology: In addition to these phases in the management of environmental risk, appropriate measures that affect the reduction of risk occurrence should be implemented: - normative and legal regulations (laws and regulations, - appropriate organizational structures in society, and - establishing quality monitoring of environment. The emphasis is placed on the application of assessment methodologies (three-model concept, as the most important aspect of successful management of environmental risk. Risk assessment methodology - European concept: The first concept of ecological risk assessment methodology is based on the so-called European model-concept.
In order to better understand this ecological risk assessment methodology, two concepts - hazard and risk - are introduced. The European concept of environmental risk assessment has the following phases in its implementation: identification of hazard (danger, identification of consequences (if there is hazard, estimate of the scale of consequences, estimate of consequence probability and risk assessment (also called risk characterization. The European concept is often used to assess risk in the environment as a model for addressing the distribution of stressors along the source - path - receptor line. Risk assessment methodology - Canadian concept: The second concept of the methodology of environmental risk assessment is based on the so-called Canadian model-concept.
The assessment of ecological risk includes risk arising from natural events (floods, extreme weather conditions, etc., technological processes and products, agents (chemical, biological, radiological, etc. 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. 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. Sung Goo Chi; Seok Jeong Park; Chul Jin Choi; Ritterbusch, S.E.; Jacob, M.C. 2002-01-01 Westinghouse Electric Company (WEC) has been working with Korea Power Engineering Company (KOPEC) on a US Department of Energy (DOE) sponsored Nuclear Energy Research Initiative (NERI) project through a collaborative agreement established for the domestic NERI program.
The project deals with Risk-Informed Assessment (RIA) of regulatory and design requirements of future nuclear power plants. An objective of the RIA project is to develop a risk-informed design process, which focuses on identifying and incorporating advanced features into future nuclear power plants (NPPs) that would meet risk goals in a cost-effective manner.
The RIA design methodology is proposed to accomplish this objective. This paper discusses the development of this methodology and demonstrates its application in the design of plant systems for future NPPs. Advanced conceptual plant systems consisting of an advanced Emergency Core Cooling System (ECCS) and Emergency Feedwater System (EFWS) for a NPP were developed and the risk-informed design process was exercised to demonstrate the viability and feasibility of the RIA design methodology. Best estimate Loss-of-Coolant Accident (LOCA) analyses were performed to validate the PSA success criteria for the NPP. The results of the analyses show that the PSA success criteria can be met using the advanced conceptual systems and that the RIA design methodology is a viable and appropriate means of designing key features of risk-significant NPP systems. (authors). Wong, S.M.; Holahan, G.M.; Chung, J.W.; Johnson, M.R.
1995-01-01 This paper describes the development and trial applications of a risk-based methodology to enhance the inspection processes for US nuclear power plants. Objectives of risk-based methods to complement prescriptive engineering approaches in US Nuclear Regulatory Commission (USNRC) inspection programs are presented. Insights from time-dependent risk profiles of plant configurational from Individual Plant Evaluation (IPE) studies were integrated to develop a framework for optimizing inspection efforts in NRC regulatory initiatives. Lessons learned from NRC pilot applications of the risk-based methodology for evaluation of the effectiveness of operational risk management programs at US nuclear power plant sites are also discussed. Matsuo, Yuji 2016-01-01 The methodology of estimating nuclear accident risk cost, as a part of nuclear power generation cost, has hardly been established due mainly to the extremely wide range of the estimation of the accident frequency. This study estimates the expected nuclear accident frequency for Japan, making use of the method of Bayesian statistics, which exploits both the information obtained by Probabilistic Risk Assessment (PRA) and the observed historical accident frequencies.
Using the PRA estimation of the Containment Failure Frequency (CFF) for Tomari nuclear power plant unit 3 of Hokkaido Electric Power Company (average: 2.1 x 10 -4, 95th percentile: 7.7 x 10 -4 ) and the actual large-scale accident frequency (once in 1,460 reactor-years), the posterior CFF was estimated at 3.8 x 10 -4. This study also took into account the 'external' factor causing unexpected nuclear accidents, concluding that such factor could result in higher CFF estimations, especially with larger observed accident numbers.
Eisinger, D.S.; Simmons, R.A.; Lammering, M.; Sotiros, R. 1991-01-01 This paper provides an easy-to-use screening methodology to estimate potential excess lifetime lung cancer risk resulting from indoor radon exposure. The methodology was developed under U.S. EPA Office of Policy, Planning, and Evaluation sponsorship of the agency's Integrated Environmental Management Projects (IEMP) and State/Regional Comparative Risk Projects. These projects help policymakers understand and use scientific data to develop environmental problem-solving strategies. This research presents the risk assessment methodology, discusses its basis, and identifies appropriate applications. The paper also identifies assumptions built into the methodology and qualitatively addresses methodological uncertainties, the direction in which these uncertainties could bias analyses, and their relative importance.
The methodology draws from several sources, including risk assessment formulations developed by the U.S. EPA's Office of Radiation Programs, the EPA's Integrated Environmental Management Project (Denver), the International Commission on Radiological Protection, and the National Institute for Occupational Safety and Health. When constructed as a spreadsheet program, the methodology easily facilitates analyses and sensitivity studies (the paper includes several sensitivity study options). The methodology will be most helpful to those who need to make decisions concerning radon testing, public education, and exposure prevention and mitigation programs.26 references.