Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors (U)

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

This paper reports on a full-scope probabilistic risk assessment (PRA) performed for the Savannah River Site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident

Probabilistic safety assessments of nuclear power plants for low power and shutdown modes

International Nuclear Information System (INIS)

2000-03-01

Within the past several years the results of nuclear power plant operating experience and performance of probabilistic safety assessments (PSAs) for low power and shutdown operating modes have revealed that the risk from operating modes other than full power may contribute significantly to the overall risk from plant operations. These early results have led to an increased focus on safety during low power and shutdown operating modes and to an increased interest of many plant operators in performing shutdown and low power PSAs. This publication was developed to provide guidance and insights on the performance of PSA for shutdown and low power operating modes. The preparation of this publication was initiated in 1994. Two technical consultants meetings were conducted in 1994 and one in February 1999 in support of the development of this report

Probabilistic analysis of 900 MWe PWR. Shutdown technical specifications

International Nuclear Information System (INIS)

Mattei, J.M.; Bars, G.

1987-11-01

During annual shutdown, preventive maintenance and modifications which are made on PWRs cause scheduled unavailabilities of equipment or systems which might harm the safety of the installation, in spite of the low level of decay heat during this period. The pumps in the auxiliary feedwater system, component cooling water system, service water system, the water injection arrays (LPIS, HPIS, CVCS), and the containment spray system may have scheduled unavailability, as well as the power supply of the electricity boards. The EDF utility is aware of the risks related to these situations for which accident procedures have been set up and hence has proposed limiting downtime for this equipment during the shutdown period, through technical specifications. The project defines the equipment required to ensure the functions important for safety during the various shutdown phases (criticality, water inventory, evacuation of decay heat, containment). In order to be able to judge the acceptability of these specifications, the IPSN, the technical support of the Service Central de Surete des Installations Nucleaires, has used probabilistic methodology to analyse the impact on the core melt probability of these specifications, for a French 900 MWe PWR

Loss of coolant accident (LOCA) analysis for McMaster Nuclear Reactor through probabilistic risk assessment (PRA)

Energy Technology Data Exchange (ETDEWEB)

Ha, T.; Garland, W.J. [McMaster Univ., Dept. of Engineering Physics, Hamilton, Ontario (Canada)]. E-mail: hats@mcmaster.ca

2006-07-01

A probabilistic risk assessment (PRA) was conducted for the loss of coolant accident (LOCA) sequence in the McMaster Nuclear Reactor (MNR). A level 1 PRA was completed including event sequence modeling, system modeling, and quantification. To support the quantification of the accident sequence identified, data analysis using the Bayesian method and human reliability analysis (HRA) using the ASEP approach were performed. Since human performance in research reactors is significantly different from that in power reactors, a different time-oriented HRA model was proposed and applied for the estimation of the human error probability (HEP) of core relocation. This HEP estimate was less than that by the ASEP approach by a factor of about 2. These two HEP estimates were used for sensitivity analysis, and modeling uncertainty in the PRA models was quantified. This showed the necessity of appropriate human reliability models in PRA for research reactors. This method could be implemented for the operators' actions which require extensive manual execution with little cognitive load, as might be the case for some maintenance operations in power reactors. (author)

Development of a methodology for conducting an integrated HRA/PRA --

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. (Brookhaven National Lab., Upton, NY (United States)); Wreathall, J. (Wreathall (John) and Co., Dublin, OH (United States)); Cooper, S.E. (Science Applications International Corp., McLean, VA (United States))

1993-01-01

During Low Power and Shutdown (LP S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP S, (2) identification of potentially important LP S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP S conditions for a pressurized water reactor (PWR).

Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

International Nuclear Information System (INIS)

Breeding, R.J.; Leahy, T.J.; Young, J.

1985-08-01

The full range of PRA topics is presented, with a special emphasis on systems analysis and PRA applications. Systems analysis topics include system modeling such as fault tree and event tree construction, failure rate data, and human Reliability. The discussion of PRA applications is centered on past and present PRA based programs, such as WASH-1400 and the Interim Reliability Evaluation Program, as well as on some of the potential future applications of PRA. The relationship of PRA to generic safety issues such as station blackout and Anticipated Transient Without Scram (ATWS) is also discussed. In addition to system modeling, the major PRA tasks of accident process analysis, and consequence analysis are presented. An explanation of the results of these activities, and the techniques by which these results are derived, forms the basis for a discussion of these topics. An additional topic which is presented in this course is the topic of PRA management, organization, and evaluation. 84 figs., 41 tabs

Results of the Level 1 probabilistic risk assessment (PRA) of internal events for heavy water production reactors

International Nuclear Information System (INIS)

Tinnes, S.P.; Cramer, D.S.; Logan, V.E.; Topp, S.V.; Smith, J.A.; Brandyberry, M.D.

1990-01-01

A full-scope probabilistic risk assessment (PRA) is being performed for the Savannah River site (SRS) production reactors. The Level 1 PRA for the K Reactor has been completed and includes the assessment of reactor systems response to accidents and estimates of the severe core melt frequency (SCMF). The internal events spectrum includes those events related directly to plant systems and safety functions for which transients or failures may initiate an accident. The SRS PRA has three principal objectives: improved understanding of SRS reactor safety issues through discovery and understanding of the mechanisms involved. Improved risk management capability through tools for assessing the safety impact of both current standard operations and proposed revisions. A quantitative measure of the risks posed by SRS reactor operation to employees and the general public, to allow comparison with declared goals and other societal risks

Outcomes of an international initiative for harmonization of low power and shutdown probabilistic safety assessment

Directory of Open Access Journals (Sweden)

Manna Giustino

2010-01-01

Full Text Available Many probabilistic safety assessment studies completed to the date have demonstrated that the risk dealing with low power and shutdown operation of nuclear power plants is often comparable with the risk of at-power operation, and the main contributors to the low power and shutdown risk often deal with human factors. Since the beginning of the nuclear power generation, human performance has been a very important factor in all phases of the plant lifecycle: design, commissioning, operation, maintenance, surveillance, modification, decommissioning and dismantling. The importance of this aspect has been confirmed by recent operating experience. This paper provides the insights and conclusions of a workshop organized in 2007 by the IAEA and the Joint Research Centre of the European Commission, on Harmonization of low power and shutdown probabilistic safety assessment for WWER nuclear power plants. The major objective of the workshop was to provide a comparison of the approaches and the results of human reliability analyses and gain insights in the enhanced handling of human factors.

Development of a methodology for conducting an integrated HRA/PRA --

International Nuclear Information System (INIS)

Luckas, W.J.; Barriere, M.T.; Brown, W.S.; Wreathall, J.; Cooper, S.E.

1993-01-01

During Low Power and Shutdown (LP ampersand S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant's systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP ampersand S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP ampersand S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP ampersand S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP ampersand S, (2) identification of potentially important LP ampersand S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP ampersand S conditions for a pressurized water reactor (PWR)

Guidance of reactor operators and TSC personnel with the severe accident management guidance under shutdown and low power conditions

International Nuclear Information System (INIS)

Van Haesendonck, M.F.; Prior, R.P.

2000-01-01

The Westinghouse Owners Group Severe Accident Management Guidance (WOG SAMG) was developed between 1991 and 1994. The primary goals for severe accident management that form the basis of the WOG SAMG are to terminate any radioactive releases to the environment; to prevent failure of any containment fission product boundary and to return the plant to a controlled stable condition. The WOG SAMG is primarily a TSC tool for mitigation of low probability core damage events. The philosophy is that control room operators should remain focused on the prevention of core damage, whereas the TSC personnel should concentrate on the mitigation of the severe accident. The symptom based package is built up as a structured process for choosing appropriate actions based on actual plant conditions. No detailed knowledge of severe accident phenomena is required. The scope of the WOG SAMG is limited to severe accidents resulting from initiating events occurring during full power operation. However, a number of studies such as the EdF EPS 1300 Probabilistic Safety Assessment (PSA), the shutdown Probabilistic Risk Assessment (PRA) for Surry, the BERA shutdown PRA for Beznau, the EPRI/ Westinghouse ORAM methodology etc. have shown that the frequency of core damage (a severe accident) during shutdown and low power operation can be of the same order of magnitude as for full power operation. The at-power SAMG is viewed as the resolution of the severe accident issue. Similarly, it is expected that as shutdown PRAs mature, the final resolution of the severe accident issue will lie in SAMG for low power and shutdown operation. Therefore in resolution of this issue, Westinghouse has developed the Shutdown Severe Accident Management Guidance (SSAMG) which gives guidance for both control room and TSC personnel to mitigate a severe accident under shutdown or low power conditions. In the last few years, many LWR plants have been implementing SAMG. In the US, all plants have developed SAMG, and many

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

International Nuclear Information System (INIS)

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

1991-01-01

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

PRA Procedures Guide: a guide to the performance of probabilistic risk assessments for nuclear power plants. Final report, Volume 1 - Chapters 1-8

International Nuclear Information System (INIS)

1983-01-01

This document, the Probabilistic Risk Assessment (PRA) Procedures Guide, is intended to provide an overview of the risk-assessment field as it exists today and to identify acceptable techniques for the systematic assessment of the risk from nuclear power plants. Topics discussed include: organization of PRA; accident-sequence definition and system modeling; human-reliability analysis; data-base development; accident-sequence quantification; physical processes of core-melt accidents; and radionuclide release and transport

PRISIM: a computer program that makes PRA useful

International Nuclear Information System (INIS)

Fussell, J.B.; Campbell, D.J.; Glynn, J.C.; Burdick, G.R.

1986-01-01

PRISIM is an IBM personal computer program that translates probabilistic risk assessment (PRA) information and calculates additional PRA type information for use by those who are not PRA experts. Specifically, PRISIM was developed for the US Nuclear Regulatory Commission for use by their resident inspectors at nuclear power plants. Inspector activities are either scheduled or are in response to a particular status of a plant. PRISIM is useful for either activity

Interaction of CREDO [Centralized Reliability Data Organization] with the EBR-II [Experimental Breeder Reactor II] PRA [probabilistic risk assessment] development

International Nuclear Information System (INIS)

Smith, M.S.; Ragland, W.A.

1989-01-01

The National Academy of Sciences review of US Department of Energy (DOE) class 1 reactors recommended that the Experimental Breeder Reactor II (EBR-II), operated by Argonne National Laboratory (ANL), develop a level 1 probabilistic risk assessment (PRA) and make provisions for level 2 and level 3 PRAs based on the results of the level 1 PRA. The PRA analysis group at ANL will utilize the Centralized Reliability Data Organization (CREDO) at Oak Ridge National Laboratory to support the PRA data needs. CREDO contains many years of empirical liquid-metal reactor component data from EBR-II. CREDO is a mutual data- and cost-sharing system sponsored by DOE and the Power Reactor and Nuclear Fuels Development Corporation of Japan. CREDO is a component based data system; data are collected on components that are liquid-metal specific, associated with a liquid-metal environment, contained in systems that interface with liquid-metal environments, or are safety related for use in reliability/availability/maintainability (RAM) analyses of advanced reactors. The links between the EBR-II PRA development effort and the CREDO data collection at EBR-II extend beyond the sharing of data. The PRA provides a measure of the relative contribution to risk of the various components. This information can be used to prioritize future CREDO data collection activities at EBR-II and other sites

Use of probabilistic risk assessment (PRA) in expert systems to advise nuclear plant operators and managers

International Nuclear Information System (INIS)

Uhrig, R.E.

1988-01-01

The use of expert systems in nuclear power plants to provide advice to managers, supervisors and/or operators is a concept that is rapidly gaining acceptance. Generally, expert systems rely on the expertise of human experts or knowledge that has been codified in publications, books, or regulations to provide advice under a wide variety of conditions. In this work, a probabilistic risk assessment (PRA) of a nuclear power plant performed previously is used to assess the safety status of nuclear power plants and to make recommendations to the plant personnel. Nuclear power plants have many redundant systems and can continue to operate when one or more of these systems is disabled or removed from service for maintenance or testing. PRAs provide a means of evaluating the risk to the public associated with the operation of nuclear power plants with components or systems out of service. While the choice of the source term and methodology in a PRA may influence the absolute probability and consequences of a core melt, the ratio of the PRA calculations for two configurations of the same plant, carried out on a consistent basis, can readily identify the increase in risk associated with going from one configuration to the other

Linkage of PRA models. Phase 1, Results

Energy Technology Data Exchange (ETDEWEB)

Smith, C.L.; Knudsen, J.K.; Kelly, D.L.

1995-12-01

The goal of the Phase I work of the ``Linkage of PRA Models`` project was to postulate methods of providing guidance for US Nuclear Regulator Commission (NRC) personnel on the selection and usage of probabilistic risk assessment (PRA) models that are best suited to the analysis they are performing. In particular, methods and associated features are provided for (a) the selection of an appropriate PRA model for a particular analysis, (b) complementary evaluation tools for the analysis, and (c) a PRA model cross-referencing method. As part of this work, three areas adjoining ``linking`` analyses to PRA models were investigated: (a) the PRA models that are currently available, (b) the various types of analyses that are performed within the NRC, and (c) the difficulty in trying to provide a ``generic`` classification scheme to groups plants based upon a particular plant attribute.

Linkage of PRA models. Phase 1, Results

International Nuclear Information System (INIS)

Smith, C.L.; Knudsen, J.K.; Kelly, D.L.

1995-12-01

The goal of the Phase I work of the ''Linkage of PRA Models'' project was to postulate methods of providing guidance for US Nuclear Regulator Commission (NRC) personnel on the selection and usage of probabilistic risk assessment (PRA) models that are best suited to the analysis they are performing. In particular, methods and associated features are provided for (a) the selection of an appropriate PRA model for a particular analysis, (b) complementary evaluation tools for the analysis, and (c) a PRA model cross-referencing method. As part of this work, three areas adjoining ''linking'' analyses to PRA models were investigated: (a) the PRA models that are currently available, (b) the various types of analyses that are performed within the NRC, and (c) the difficulty in trying to provide a ''generic'' classification scheme to groups plants based upon a particular plant attribute

Standardized procedure for tsunami PRA by AESJ

International Nuclear Information System (INIS)

Kirimoto, Yukihiro; Yamaguchi, Akira; Ebisawa, Katsumi

2013-01-01

After Fukushima Accident (March 11, 2011), the Atomic Energy Society of Japan (AESJ) started to develop the standard of Tsunami Probabilistic Risk Assessment (PRA) for nuclear power plants in May 2011. As Japan is one of the countries with frequent earthquakes, a great deal of efforts has been made in the field of seismic research since the early stage. To our regret, the PRA procedures guide for tsunami has not yet been developed although the importance is held in mind of the PRA community. Accordingly, AESJ established a standard to specify the standardized procedure for tsunami PRA considering the results of investigation into the concept, the requirements that should have and the concrete methods regarding tsunami PRA referring the opinions of experts in the associated fields in December 2011 (AESJ-SC-RK004:2011). (author)

PRA-Code Upgrade to Handle a Generic Problem

International Nuclear Information System (INIS)

Wilson, J. R.

1999-01-01

During the probabilistic risk assessment (PRA) for the proposed Yucca Mountain nuclear waste repository, a problem came up that could not be handled by most PRA computer codes. This problem deals with dependencies between sequential events in time. Two similar scenarios that illustrate this problem are LOOP nonrecovery and sequential wearout failures with units of time. The purpose of this paper is twofold: To explain the problem generically, and to show how the PRA code at the INEEL, SAPHIRE, has been modified to solve this problem correctly

Summary of PRA assessment of transient accident risks, human factors considerations, and PRA methods and applications

International Nuclear Information System (INIS)

Carnino, A.

1984-01-01

This chapter reviews the progress made in the probabilistic risk assessment (PRA) area to help in solving operational transient problems and to integrate human factors considerations, as discussed at the American Nuclear Society Topical Meeting on Anticipated and Abnormal Plant Transients in Light Water Reactors. Topics considered include core-melt frequency, external events (e.g., fires, floods), diagnostic errors, and operator aids. It is concluded that confidence in PRA results, predictions and uses for decisions in both the safety of the plants and their availability will improve

Use of probabilistic risk assessment (PRA) in expert systems to advise nuclear plant operators and managers

International Nuclear Information System (INIS)

Uhrig, R.E.

1988-01-01

The use of expert systems in nuclear power plants to provide advice to managers, supervisors and/or operators is a concept that is rapidly gaining acceptance. Generally, expert systems rely on the expertise of human experts or knowledge that has been modified in publications, books, or regulations to provide advice under a wide variety of conditions. In this work, a probabilistic risk assessment (PRA) 3 of a nuclear power plant performed previously is used to assess the safety status of nuclear power plants and to make recommendations to the plant personnel. 5 refs., 1 fig., 2 tabs

Insights on PRA Review Practices: Necessity for Model Shaking

Energy Technology Data Exchange (ETDEWEB)

Kim, Inn Seock; Jang, Mi suk; Kim, Seoung Rae [NESS, Daejeon (Korea, Republic of)

2016-05-15

Probabilistic risk assessment (PRA) is increasingly used as a technique to help ensure design and operational safety of nuclear power plants (NPPs) in the nuclear industry. Hence, there is considerable interest in the PRA quality, and as a result, a peer review of the PRA model is typically performed to ensure its technical adequacy as part of the PRA development process or for any other reason (e.g., regulatory requirement). For the PRA model to be used as a valuable vehicle for risk-informed applications, it is essential that the PRA model must yield correct and physically meaningful accident sequences and minimal cutsets for specific plant configurations or conditions relating to the applications. Hence, the existing peer review guidelines need to be updated to reflect these insights so that risk-informed applications could be more actively pursued with confidence.

Review of PRA methodology for LMFBR

International Nuclear Information System (INIS)

Yang, J. E.

1999-02-01

Probabilistic Risk Assessment (PRA) has been widely used as a tool to evaluate the safety of NPPs (Nuclear Power Plants), which are in the design stage as well as in operation. Recently, PRA becomes one of the licensing requirements for many existing and new NPPs. KALIMER is a Liquid Metal Fast Breeder Reactor (LMFBR) being developed by KAERI. Since the design concept of KALIMER is similar to that of the PRISM plant developed by GE, it would be appropriate to review the PRA methodology of PRISM as the first step of KALIMER PRA. Hence, in this report summarizes the PRA methodology of PRISM plant, and the required works for the PSA of KALIMER based on the reviewed results. The PRA technology of PRISM plant consists of following five major tasks: (1) development of initiating event list, (2) development of system event tree, (3) development of core response event tree, (4) development of containment response event tree, and (5) consequences and risk estimation. The estimated individual and societal risk measures show that the risk from a PRISM module is substantially less than the NRC goal. Each task is compared to the PRA methodology of Light Water Reactor (LWR)/Pressurized Heavy Water Reactor (PHWR). In the report, each task of PRISM PRA methodology is reviewed and compared to the corresponding part of LWR/PHWR PSA performed in Korea. The parts that are not modeled appropriately in PRISM PRA are identified, and the recommendations for KALIMER PRA are stated. (author). 14 refs., 9 tabs., 4 figs

How the chemical industry can benefit from PRA

International Nuclear Information System (INIS)

Guymer, P.; Kaiser, G.D.; Mc Kelvey, T.W.; Hannaman, G.W.

1986-01-01

Probabilistic Risk Assessment (PRA) is a method of quantifying the frequency of occurrence and the magnitude of the consequences of accidents in systems that contain hazardous materials such as radioactive fission products, and toxic, flammable or explosive chemicals. The frequency and the magnitude of the consequences are the basic elements of any definition or risk, which is often simply expressed as the product of frequency and magnitude, summed over all accident sequences. PRA is now a mature technique that has been used to estimate risk for a number of industrial facilities. In this paper the author gives examples of beneficial uses of PRA

Role of seismic PRA in seismic safety decisions of nuclear power plants

International Nuclear Information System (INIS)

Ravindra, M.K.; Kennedy, R.P.; Sues, R.H.

1985-01-01

This paper highlights the important roles that seismic probabilistic risk assessments (PRAs) can play in the seismic safety decisions of nuclear power plants. If a seismic PRA has been performed for a plant, its results can be utilized to evaluate the seismic capability beyond the safe shutdown event (SSE). Seismic fragilities of key structures and equipment, fragilities of dominant plant damage states and the frequencies of occurrence of these plant damage states are reviewed to establish the seismic safety of the plant beyond the SSE level. Guidelines for seismic margin reviews and upgrading may be developed by first identifying the generic classes of structures and equipment that have been shown to be dominant risk contributors in the completed seismic PRAs, studying the underlying causes for their contribution and examining why certain other items (e.g., piping) have not proved to be high-risk-contributors

Documentation design for probabilistic risk assessment

International Nuclear Information System (INIS)

Parkinson, W.J.; von Herrmann, J.L.

1985-01-01

This paper describes a framework for documentation design of probabilistic risk assessment (PRA) and is based on the EPRI document NP-3470 ''Documentation Design for Probabilistic Risk Assessment''. The goals for PRA documentation are stated. Four audiences are identified which PRA documentation must satisfy, and the documentation consistent with the needs of the various audiences are discussed, i.e., the Summary Report, the Executive Summary, the Main Report, and Appendices. The authors recommend the documentation specifications discussed herein as guides rather than rigid definitions

Probabilistic Reversible Automata and Quantum Automata

OpenAIRE

Golovkins, Marats; Kravtsev, Maksim

2002-01-01

To study relationship between quantum finite automata and probabilistic finite automata, we introduce a notion of probabilistic reversible automata (PRA, or doubly stochastic automata). We find that there is a strong relationship between different possible models of PRA and corresponding models of quantum finite automata. We also propose a classification of reversible finite 1-way automata.

Current status and future expectation concerning probabilistic risk assessment of NPPs. 1. Features and issues of probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Yamashita, Masahiro

2012-01-01

Probabilistic risk assessment (PRA) of Nuclear Power Plants (NPPs) could play an important role in assuring safety of NPPs. However PRA had not always effectively used, which was indicated in Japanese government's report on Fukushima Daiichi NPP accident. At the Risk Technical Committee (RTC) of Standards Committee of Atomic Energy Society of Japan, preparation of standards (implementing criteria) focusing on PRA methodology and investigation on basic philosophy for use of PRA had been in progress. Based on activities of RTC, a serial in three articles including this described current status and future expectation concerning probabilistic risk assessment of NPPs. This article introduced features and issues of PRA methodology related to the use of PRA. Features of PRA methodology could be shown as (1) systematic and comprehensive understanding of risk, (2) support of grading approach, (3) identification of effective safety upgrade measures and (4) quantitative understanding of effects of uncertainty. Issues of PRA methodology were (1) extension of PRA application area, (2) upgrade of PRA methodology, (3) quality assurance of PRA, (4) treatment of uncertainty and (5) quantitative evaluation criteria. (T. Tanaka)

PRA and Risk Informed Analysis

International Nuclear Information System (INIS)

Bernsen, Sidney A.; Simonen, Fredric A.; Balkey, Kenneth R.

2006-01-01

The Boiler and Pressure Vessel Code (BPVC) of the American Society of Mechanical Engineers (ASME) has introduced a risk based approach into Section XI that covers Rules for Inservice Inspection of Nuclear Power Plant Components. The risk based approach requires application of the probabilistic risk assessments (PRA). Because no industry consensus standard existed for PRAs, ASME has developed a standard to evaluate the quality level of an available PRA needed to support a given risk based application. The paper describes the PRA standard, Section XI application of PRAs, and plans for broader applications of PRAs to other ASME nuclear codes and standards. The paper addresses several specific topics of interest to Section XI. Important consideration are special methods (surrogate components) used to overcome the lack of PRA treatments of passive components in PRAs. The approach allows calculations of conditional core damage probabilities both for component failures that cause initiating events and failures in standby systems that decrease the availability of these systems. The paper relates the explicit risk based methods of the new Section XI code cases to the implicit consideration of risk used in the development of Section XI. Other topics include the needed interactions of ISI engineers, plant operating staff, PRA specialists, and members of expert panels that review the risk based programs

Practical PRA applications at Consumers Power Company

International Nuclear Information System (INIS)

Blanchard, D.P.

1985-01-01

Consumers Power Company has completed two probabilistic risk assessments (PRAs), one each at its Big Rock Point and Midland plants and is in the process of performing a third study at its Palisades Plant. Each PRA is summarized briefly in this paper. Each PRA has been used to evaluate specific plant design features and make operating and design recommendations to plant and Company management as well as to the regulator. This paper is a sumary of those issues on which Consumers Power Company has applied PRAs to date. The technique used in applying PRA to these issues has varied as more was learned about the plants from the PRA and about PRA itself. Some issue resolutions involved deriving technical arguments from small parts of the PRA only, such as the logic models or consequence analysis. Still others required use of the entire PRA including sequence quantification, plant and containment response, consequence analysis and eventually cost-benefit evaluation of proposed resolutions. The benefits derived from these analyses have also varied and include not only a perceived reduction in the risks associated with plant operation but also economic benefit to the Company in that cost-effective alternatives to resolving safety issues have been permitted

PRA: A Perspective on Strengths, Current Limitations, And Possible Improvements

International Nuclear Information System (INIS)

Mosleh, Ail

2014-01-01

Probabilistic risk assessment (PRA) has been used in various technological fields to assist regulatory agencies, managerial decision makers, and systems designers in assessing and mitigating the risks inherent in these complex arrangements. Has PRA delivered on its promise? How do we gage PRA performance? Are our expectations about value of PRA realistic? Are there disparities between what we get and what we think we are getting form PRA and its various derivatives? Do current PRAs reflect the knowledge gained from actual events? How do we address potential gaps? These are some of the questions that have been raised over the years since the inception of the field more than forty years ago. This paper offers a brief assessment of PRA as a technical discipline in theory and practice, its key strengths and weaknesses, and suggestions on ways to address real and perceived shortcomings

External event Probabilistic Risk Assessment for the High Flux Isotope Reactor (HFIR)

International Nuclear Information System (INIS)

Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H.

1989-01-01

The High Flux Isotope Reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988 a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 x 10 -4 . In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 50% of the internal event initiated contribution and is dominated by seismic events

Chinshan living PRA model using NUPRA software package

International Nuclear Information System (INIS)

Cheng, S.-K.; Lin, T.-J.

2004-01-01

A living probabilistic risk assessment (PRA) model has been established for Chinshan Nuclear Power Station (BWR-4, MARK-I) using NUPRA software package. The core damage frequency due to internal events, seismic events and typhoons are evaluated in this model. The methodology and results considering the recent implementation of the 5th emergency diesel generator and automatic boron injection function are presented. The dominant sequences of this PRA model are discussed, and some possible applications of this living model are proposed. (author)

PRA: A PERSPECTIVE ON STRENGTHS, CURRENT LIMITATIONS, AND POSSIBLE IMPROVEMENTS

Directory of Open Access Journals (Sweden)

ALI MOSLEH

2014-02-01

Full Text Available Probabilistic risk assessment (PRA has been used in various technological fields to assist regulatory agencies, managerial decision makers, and systems designers in assessing and mitigating the risks inherent in these complex arrangements. Has PRA delivered on its promise? How do we gage PRA performance? Are our expectations about value of PRA realistic? Are there disparities between what we get and what we think we are getting form PRA and its various derivatives? Do current PRAs reflect the knowledge gained from actual events? How do we address potential gaps? These are some of the questions that have been raised over the years since the inception of the field more than forty years ago. This paper offers a brief assessment of PRA as a technical discipline in theory and practice, its key strengths and weaknesses, and suggestions on ways to address real and perceived shortcomings.

Development of insights from PRAs for non-PRA people

International Nuclear Information System (INIS)

Reilly, H.J.; Meale, B.M.

1992-01-01

A probabilistic risk assessment (PRA) of the Savannah River K-Reactor was completed in 1990. The PRA estimated the frequency of core damage accidents caused by operational occurrences during power operation of the reactor. The US Department of Energy (DOE) requested Idaho National Engineering Laboratory (INEL) to prepare guidance based on the PRA for use by DOE personnel at the Savannah River Site (SRS). The document had the purpose of informing the DOE system engineers and site representatives about how the information in the PRA might be used to help guide their activities. Opportunities existed to develop a document somewhat different than those developed previously by other programs. The opportunities existed because the audience is different: the principal audience for the document consists of DOE engineers who have continuing oversight responsibility for activities performed by the operating contractor at the K-Reactor, but who may not be knowledgeable about PRA

Recovery actions in PRA [probabilistic risk assessment] for the Risk Methods Integration and Evaluation Program (RMIEP): Volume 1, Development of the data-based method

International Nuclear Information System (INIS)

Weston, L.M.; Whitehead, D.W.; Graves, N.L.

1987-06-01

In a probabilistic risk assessment (PRA) for a nuclear power plant, the analyst identifies a set of potential core damage events consisting of equipment failures and human errors and their estimated probabilities of occurrence. If operator recovery from an event within some specified time is considered, then the probability of this recovery can be included in the PRA. This report provides PRA analysts with an improved methodology for including recovery actions in a PRA. A recovery action can be divided into two distinct phases: a Diagnosis Phase (realizing that there is a problem with a critical parameter and deciding upon the correct course of action) and an Action Phase (physically accomplishing the required action). In this methodology, simulator data are used to estimate recovery probabilities for the diagnosis phase. Different time-reliability curves showing the probability of failure of diagnosis as a function of time from the compelling cue for the event are presented. These curves are based on simulator exercises, and the actions are grouped based upon their operational similarities. This is an improvement over existing diagnosis models that rely greatly upon subjective judgment to obtain such estimates. The action phase is modeled using estimates from available sources. The methodology also includes a recommendation on where and when to apply the recovery action in the PRA process

The tsunami probabilistic risk assessment (PRA). Example of accident sequence analysis of tsunami PRA according to the standard for procedure of tsunami PRA for nuclear power plants

International Nuclear Information System (INIS)

Ohara, Norihiro; Hasegawa, Keiko; Kuroiwa, Katsuya

2013-01-01

After the Fukushima Daiichi nuclear power plant (NPP) accident, standard for procedure of tsunami PRA for NPP had been established by the Standardization Committee of AESJ. Industry group had been conducting analysis of Tsunami PRA for PWR based on the standard under the cooperation with electric utilities. This article introduced overview of the standard and examples of accident sequence analysis of Tsunami PRA studied by the industry group according to the standard. The standard consisted of (1) investigation of NPP's composition, characteristics and site information, (2) selection of relevant components for Tsunami PRA and initiating events and identification of accident sequence, (3) evaluation of Tsunami hazards, (4) fragility evaluation of building and components and (5) evaluation of accident sequence. Based on the evaluation, countermeasures for further improvement of safety against Tsunami could be identified by the sensitivity analysis. (T. Tanaka)

Use of limited data to construct Bayesian networks for probabilistic risk assessment.

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina M.; Swiler, Laura Painton

2013-03-01

Probabilistic Risk Assessment (PRA) is a fundamental part of safety/quality assurance for nuclear power and nuclear weapons. Traditional PRA very effectively models complex hardware system risks using binary probabilistic models. However, traditional PRA models are not flexible enough to accommodate non-binary soft-causal factors, such as digital instrumentation&control, passive components, aging, common cause failure, and human errors. Bayesian Networks offer the opportunity to incorporate these risks into the PRA framework. This report describes the results of an early career LDRD project titled %E2%80%9CUse of Limited Data to Construct Bayesian Networks for Probabilistic Risk Assessment%E2%80%9D. The goal of the work was to establish the capability to develop Bayesian Networks from sparse data, and to demonstrate this capability by producing a data-informed Bayesian Network for use in Human Reliability Analysis (HRA) as part of nuclear power plant Probabilistic Risk Assessment (PRA). This report summarizes the research goal and major products of the research.

Certification plan for safety and PRA codes

International Nuclear Information System (INIS)

Toffer, H.; Crowe, R.D.; Ades, M.J.

1990-05-01

A certification plan for computer codes used in Safety Analyses and Probabilistic Risk Assessment (PRA) for the operation of the Savannah River Site (SRS) reactors has been prepared. An action matrix, checklists, and a time schedule have been included in the plan. These items identify what is required to achieve certification of the codes. A list of Safety Analysis and Probabilistic Risk Assessment (SA ampersand PRA) computer codes covered by the certification plan has been assembled. A description of each of the codes was provided in Reference 4. The action matrix for the configuration control plan identifies code specific requirements that need to be met to achieve the certification plan's objectives. The checklist covers the specific procedures that are required to support the configuration control effort and supplement the software life cycle procedures based on QAP 20-1 (Reference 7). A qualification checklist for users establishes the minimum prerequisites and training for achieving levels of proficiency in using configuration controlled codes for critical parameter calculations

Level 2 PRA for a German BWR

International Nuclear Information System (INIS)

Sassen, F.; Rapp, W.; Tietsch, W.; Roess, P.

2007-01-01

A concept for a Level 2 Probabilistic Risk Assessment (L2 PRA) for a German Boiling Water Reactor (BWR) has been developed taking into account the role of L2 PRA within the German regulatory landscape. According to this concept, a plant specific evaluation of the severe accident phenomenology as well as analyses of the accident progression for the severe accident scenarios has been performed. Furthermore a plant specific MELCOR 1.8.6 model has been developed and special MELCOR source term calculations have been performed for the different release paths. This paper will present examples from the different areas described above. (author)

Development of a methodology for conducting an integrated HRA/PRA --. Task 1, An assessment of human reliability influences during LP&S conditions PWRs

Energy Technology Data Exchange (ETDEWEB)

Luckas, W.J.; Barriere, M.T.; Brown, W.S. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., McLean, VA (United States)

1993-06-01

During Low Power and Shutdown (LP&S) conditions in a nuclear power plant (i.e., when the reactor is subcritical or at less than 10--15% power), human interactions with the plant`s systems will be more frequent and more direct. Control is typically not mediated by automation, and there are fewer protective systems available. Therefore, an assessment of LP&S related risk should include a greater emphasis on human reliability than such an assessment made for power operation conditions. In order to properly account for the increase in human interaction and thus be able to perform a probabilistic risk assessment (PRA) applicable to operations during LP&S, it is important that a comprehensive human reliability assessment (HRA) methodology be developed and integrated into the LP&S PRA. The tasks comprising the comprehensive HRA methodology development are as follows: (1) identification of the human reliability related influences and associated human actions during LP&S, (2) identification of potentially important LP&S related human actions and appropriate HRA framework and quantification methods, and (3) incorporation and coordination of methodology development with other integrated PRA/HRA efforts. This paper describes the first task, i.e., the assessment of human reliability influences and any associated human actions during LP&S conditions for a pressurized water reactor (PWR).

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Main report and appendices, Volume 6, Part 1

International Nuclear Information System (INIS)

Brown, T.D.; Kmetyk, L.N.; Whitehead, D.; Miller, L.; Forester, J.; Johnson, J.

1995-03-01

Traditionally, probabilistic risk assessments (PRAS) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Recent studies and operational experience have, however, implied that accidents during low power and shutdown could be significant contributors to risk. In response to this concern, in 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The program consists of two parallel projects being performed by Brookhaven National Laboratory (Surry) and Sandia National Laboratories (Grand Gulf). The program objectives include assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing the estimated risks with the risk associated with accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program is that of a Level-3 PRA. The subject of this report is the PRA of the Grand Gulf Nuclear Station, Unit 1. The Grand Gulf plant utilizes a 3833 MWt BUR-6 boiling water reactor housed in a Mark III containment. The Grand Gulf plant is located near Port Gibson, Mississippi. The regime of shutdown analyzed in this study was plant operational state (POS) 5 during a refueling outage, which is approximately Cold Shutdown as defined by Grand Gulf Technical Specifications. The entire PRA of POS 5 is documented in a multi-volume NUREG report (NUREG/CR-6143). The internal events accident sequence analysis (Level 1) is documented in Volume 2. The Level 1 internal fire and internal flood analyses are documented in Vols 3 and 4, respectively

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Main report and appendices, Volume 6, Part 1

Energy Technology Data Exchange (ETDEWEB)

Brown, T.D.; Kmetyk, L.N.; Whitehead, D.; Miller, L. [Sandia National Labs., Albuquerque, NM (United States); Forester, J. [Science Applications International Corp., Albuquerque, NM (United States); Johnson, J. [GRAM, Inc., Albuquerque, NM (United States)

1995-03-01

Traditionally, probabilistic risk assessments (PRAS) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Recent studies and operational experience have, however, implied that accidents during low power and shutdown could be significant contributors to risk. In response to this concern, in 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The program consists of two parallel projects being performed by Brookhaven National Laboratory (Surry) and Sandia National Laboratories (Grand Gulf). The program objectives include assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing the estimated risks with the risk associated with accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program is that of a Level-3 PRA. The subject of this report is the PRA of the Grand Gulf Nuclear Station, Unit 1. The Grand Gulf plant utilizes a 3833 MWt BUR-6 boiling water reactor housed in a Mark III containment. The Grand Gulf plant is located near Port Gibson, Mississippi. The regime of shutdown analyzed in this study was plant operational state (POS) 5 during a refueling outage, which is approximately Cold Shutdown as defined by Grand Gulf Technical Specifications. The entire PRA of POS 5 is documented in a multi-volume NUREG report (NUREG/CR-6143). The internal events accident sequence analysis (Level 1) is documented in Volume 2. The Level 1 internal fire and internal flood analyses are documented in Vols 3 and 4, respectively.

Selecting the seismic HRA approach for Savannah River Plant PRA revision 1

International Nuclear Information System (INIS)

Papouchado, K.; Salaymeh, J.

1993-10-01

The Westinghouse Savannah River Company (WSRC) has prepared a level I probabilistic risk assessment (PRA), Rev. 0 of reactor operations for externally-initiated events including seismic events. The SRS PRA, Rev. 0 Seismic HRA received a critical review that expressed skepticism with the approach used for human reliability analysis because it had not been previously used and accepted in other published PRAs. This report provides a review of published probabilistic risk assessments (PRAs), the associated methodology guidance documents, and the psychological literature to identify parameters important to seismic human reliability analysis (HRA). It also describes a recommended approach for use in the Savannah River Site (SRS) PRA. The SRS seismic event PRA performs HRA to account for the contribution of human errors in the accident sequences. The HRA of human actions during and after a seismic event is an area subject to many uncertainties and involves significant analyst judgment. The approach recommended by this report is based on seismic HRA methods and associated issues and concerns identified from the review of these referenced documents that represent the current state-of-the- art knowledge and acceptance in the seismic HRA field

PRA: a powerful engineering decision tool

International Nuclear Information System (INIS)

Carvalho, H.G. de.

1988-03-01

The probabilistic risk analysis (PRA) is studied and its historical development is briefly presented. Human factors, sofware and guides, improvement of utility management of nuclear power operations are discussed. The development of a standardized LWR design, optimized for safety, reliability and economy is studied. The impact of risk assessments in public acceptance of nuclear power is discussed. (M.A.C.) [pt

Risk contribution from low power, shutdown, and other operational modes beyond full power

Energy Technology Data Exchange (ETDEWEB)

Whitehead, D.W.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States); Chu, T.L. [Brookhaven National Lab., Upton, NY (United States)] [and others

1995-04-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 probabilistic risk assessment (PRA) for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. A phased approach was used in Level 1. In Phase 1 the concept of plant operational states (POSs) was developed to provide a better representation of the plant as it transitions from power to nonpower operation. This included a coarse screening analysis of all POSs to identify vulnerable plant configurations, to characterize (on a high, medium, or low basis) potential frequencies of core damage accidents, and to provide a foundation for a detailed Phase 2 analysis. In Phase 2, selected POSs from both Grand Gulf and Surry were chosen for detailed analysis. For Grand Gulf, POS 5 (approximately cold shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected. For Surry, three POSs representing the time the plant spends in midloop operation were chosen for analysis. These included POS 6 and POS 10 of a refueling outage and POS 6 of a drained maintenance outage. Level 1 and Level 2/3 results from both the Surry and Grand Gulf analyses are presented.

Risk contribution from low power, shutdown, and other operational modes beyond full power

International Nuclear Information System (INIS)

Whitehead, D.W.; Brown, T.D.; Chu, T.L.; Pratt, W.T.

1995-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 probabilistic risk assessment (PRA) for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. A phased approach was used in Level 1. In Phase 1 the concept of plant operational states (POSs) was developed to provide a better representation of the plant as it transitions from power to nonpower operation. This included a coarse screening analysis of all POSs to identify vulnerable plant configurations, to characterize (on a high, medium, or low basis) potential frequencies of core damage accidents, and to provide a foundation for a detailed Phase 2 analysis. In Phase 2, selected POSs from both Grand Gulf and Surry were chosen for detailed analysis. For Grand Gulf, POS 5 (approximately cold shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected. For Surry, three POSs representing the time the plant spends in midloop operation were chosen for analysis. These included POS 6 and POS 10 of a refueling outage and POS 6 of a drained maintenance outage. Level 1 and Level 2/3 results from both the Surry and Grand Gulf analyses are presented

Role of PRA in new NPP projects

International Nuclear Information System (INIS)

Julin, A.; Sandberg, J.; Virolainen, R.

2012-01-01

In Finland, a plant specific, Level 1 and 2 Probabilistic Risk Analysis (PRA) is required as a prerequisite for issuing the construction license and operating license. The use of PRA in various applications and the main insights are presented. These applications include e.g. PRA support to the design of SSCs (Systems, Structures and Components), definition of pre-service and in-service inspection programs, evaluation of the safety classification of SSCs, development of procedures, training and in definition of risk informed technical specifications, periodic testing and on-line preventive maintenance programs. In addition, PRA shall be used to assess the adequacy and coverage of the phase and system commissioning programs. Also the potential risks related to commissioning tests during nuclear test phase, shall be assessed with the help of PRA. In OL3 project, risk informed approach has been applied on a large scale for the first time in the design, construction and commissioning of a new NPP unit. Pre-nuclear commissioning tests have started at OL3 site and the plant is foreseen to begin commercial operation in 2013. Decisions have been made to launch new NPP projects. Teollisuuden Voima Oyj (TVO) is planning to build a new unit (OL4) at Olkiluoto site and a new utility, Fennovoima, is planning to build one unit at one of two alternative green field sites in Northern parts of Finland. Insights from PRAs of operating NPPs have been used in the evaluation of possible new sites to ensure that the site specific concerns and environmental conditions are adequately taken into account in the design of SSCs. Although the seismic activity at the Olkiluoto site is low, a comprehensive seismic risk analysis is being conducted. Its results support the review of the deterministic seismic design. For new sites, a probabilistic seismic hazard analysis has been carried out for the determination of the design earthquake. Experiences from OL3 licensing have been utilized in the

Probabilistic Risk Assessment to Inform Decision Making: Frequently Asked Questions

Science.gov (United States)

General concepts and principles of Probabilistic Risk Assessment (PRA), describe how PRA can improve the bases of Agency decisions, and provide illustrations of how PRA has been used in risk estimation and in describing the uncertainty in decision making.

Reliability design of a critical facility: An application of PRA methods

International Nuclear Information System (INIS)

Souza Vieira Neto, A.; Souza Borges, W. de

1987-01-01

Although a general agreement concerning the enforcement of reliability (probabilistic) design criteria for nuclear utilities is yet to be achieved. PRA methodology can still be used successfully as a project design and review tool, aimed at improving system's prospective performance or minimizing expected accident consequences. In this paper, the potential of such an application of PRA methods is examined in the special case of a critical design project currently being developed in Brazil. (orig.)

Implications of probabilistic risk assessment

International Nuclear Information System (INIS)

Cullingford, M.C.; Shah, S.M.; Gittus, J.H.

1987-01-01

Probabilistic risk assessment (PRA) is an analytical process that quantifies the likelihoods, consequences and associated uncertainties of the potential outcomes of postulated events. Starting with planned or normal operation, probabilistic risk assessment covers a wide range of potential accidents and considers the whole plant and the interactions of systems and human actions. Probabilistic risk assessment can be applied in safety decisions in design, licensing and operation of industrial facilities, particularly nuclear power plants. The proceedings include a review of PRA procedures, methods and technical issues in treating uncertainties, operating and licensing issues and future trends. Risk assessment for specific reactor types or components and specific risks (eg aircraft crashing onto a reactor) are used to illustrate the points raised. All 52 articles are indexed separately. (U.K.)

Overview of the probabilistic risk assessment approach

International Nuclear Information System (INIS)

Reed, J.W.

1985-01-01

The techniques of probabilistic risk assessment (PRA) are applicable to Department of Energy facilities. The background and techniques of PRA are given with special attention to seismic, wind and flooding external events. A specific application to seismic events is provided to demonstrate the method. However, the PRA framework is applicable also to wind and external flooding. 3 references, 8 figures, 1 table

Incorporating organizational factors into Probabilistic Risk Assessment (PRA) of complex socio-technical systems: A hybrid technique formalization

International Nuclear Information System (INIS)

Mohaghegh, Zahra; Kazemi, Reza; Mosleh, Ali

2009-01-01

This paper is a result of a research with the primary purpose of extending Probabilistic Risk Assessment (PRA) modeling frameworks to include the effects of organizational factors as the deeper, more fundamental causes of accidents and incidents. There have been significant improvements in the sophistication of quantitative methods of safety and risk assessment, but the progress on techniques most suitable for organizational safety risk frameworks has been limited. The focus of this paper is on the choice of 'representational schemes' and 'techniques.' A methodology for selecting appropriate candidate techniques and their integration in the form of a 'hybrid' approach is proposed. Then an example is given through an integration of System Dynamics (SD), Bayesian Belief Network (BBN), Event Sequence Diagram (ESD), and Fault Tree (FT) in order to demonstrate the feasibility and value of hybrid techniques. The proposed hybrid approach integrates deterministic and probabilistic modeling perspectives, and provides a flexible risk management tool for complex socio-technical systems. An application of the hybrid technique is provided in the aviation safety domain, focusing on airline maintenance systems. The example demonstrates how the hybrid method can be used to analyze the dynamic effects of organizational factors on system risk

Issues and insights of PRA methodology in nuclear and space applications

International Nuclear Information System (INIS)

Hsu, F.

2005-01-01

This paper presents some important issues and technical insights on the scope, conceptual framework, and essential elements of nuclear power plant Probabilistic Risk Assessments (PRAs) and that of the PRAs in general applications of the aerospace industry, such as the Space Shuttle PRA being conducted by NASA. Discussions are focused on various lessons learned in nuclear power plant PRA applications and their potential applicability to the PRAs in the aerospace and launch vehicle systems. Based on insights gained from PRA projects for nuclear power plants and from the current Space Shuttle PRA effort, the paper explores the commonalities and the differences between the conduct of the different PRAs and the key issues and risk insights derived from extensive modeling practices in both industries of nuclear and space. (author)

Uses of PRA in nuclear reactor regulation

International Nuclear Information System (INIS)

Congel, F.

1987-01-01

For the past five years, more than ten probabilistic risk assessment (PRA) studies were conducted by the owners of nuclear utilities and were submitted for the review of US Nuclear Regulatory Commission staff. These PRA studies were reviewed under various types of regulatory activities depending on the nature of plant licensing stage. The reviews of these PRAs provided very valuable uses to both the staff and the licensees on safety matters of the plant operation. The licensees developed perspectives using PRA models on the safety profiles of their plants. These PRA perspectives influenced licensees' major decisions to implement improvements to plant design and operating and emergency procedures to reduce and/or eliminate the plant's vulnerability to core damage accidents. The staff's review of these PRAs particularly emphasized the dominant accident sequences. The resulting findings led to the identification of dominant risk contributors, critical areas of plant locations, mechanisms leading to potential early containment failures, and instances of noncompliances of staff's deterministic criteria. Specific examples include single failure criterion and separation requirements to assess the need for any additional measures to further improve the safety of the plant. Some of these PRAs were reviewed under regulatory activities other than safety review such as environmental review, final design review, and licensing hearings. Most importantly, the risk profiles of generic PRAs will continue to be used in reviewing and evaluating unresolved safety issues and other generic issues. The major regulatory uses of PRAs, a summary of full scope PRA review, a summary of plant improvements as a result of PRA reviews, and the future role of PRA reviews are presented

'Living PRA' concept for plant risk: Reliability and availability tracking

International Nuclear Information System (INIS)

Sancaktar, S.; Sharp, D.R.

1985-01-01

The 'Living PRA' (Probabilistic Risk Assessment) is based on placing a PRA plant model on an interactive computer. This model consists of fault tree analyses for plant systems, event tree analyses for abnormal events and site specific consequence analysis for public and/or financial risks, for a nuclear power plant. A living PRA allows updates and sensitivity analyses by the plant owner throughout the lifetime of a plant. Recently, event and fault trees from two major PRAs were placed in a computerized format. The BYRON PRA study and the Living PRA and Economic Risk examples for Indian Point Unit-3 enabled analysts to gain experience and insight into the problems of plant operation. The above concept is well established for the Nuclear Power Plant evaluation. It has been also used for evaluation of processing facilities. In these studies, systems modeling was carried out by using the GRAFTER system for automated fault tree construction. Presently both the tools and the experience exists to set up useful and viable living PRA models for nuclear and chemical processing plants to enhance risk management by the plant owners through in-house use of micro computer based models

Insights gained through probabilistic risk assessments

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven probabilistic risk assessments (PRA) studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants

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)

The Angra 1 fire PRA project

International Nuclear Information System (INIS)

Silva, Luiz E. Massiere de C.; Kassawara, Robert

2009-01-01

The Angra 1 Fire PRA (Probabilistic Risk Assessment) is under development by ELETRONUCLEAR jointly with EPRI (Electric Power Research Institute). The project was started January of 2007 and it is foreseen to be finished in the middle of the next year. The study is being conducted according to the newest methodology developed by EPRI and NRC/RES (U.S. Nuclear Regulatory Commission - Office of Regulatory Research) published in 2005 as Fire PRA Methodology for Nuclear Power Facilities (NUREG/CR-6850 or EPRI TR-1011989) [1]. Starting from the Internal Events Angra 1 PRA model Level 1 the project aims to be a comprehensive plant-specific fire analysis to identify the possible consequences of a fire in the plant vital areas which threaten the integrity of systems relevant to the safety, challenging the safety functions and representing a risk of accident that can lead to a core damage. The main tasks include the plant boundary and partitioning, the fire PRA component selection and the identification of the possible fire scenarios (ignition, propagation, detection, extinction and hazards) considering human failure events to establish the fire-induced risk model for quantification of the risk for nuclear core damage taking into account the plant design and its fire protection resources. This work presents a general discussion on the methodology applied to the completed steps of the project. (author)

How Can You Support RIDM/CRM/RM Through the Use of PRA

Science.gov (United States)

DoVemto. Tpmu

2011-01-01

Probabilistic Risk Assessment (PRA) is one of key Risk Informed Decision Making (RIDM) tools. It is a scenario-based methodology aimed at identifying and assessing Safety and Technical Performance risks in complex technological systems.

Incorporating organizational factors into Probabilistic Risk Assessment (PRA) of complex socio-technical systems: A hybrid technique formalization

Energy Technology Data Exchange (ETDEWEB)

Mohaghegh, Zahra [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States)], E-mail: mohagheg@umd.edu; Kazemi, Reza; Mosleh, Ali [Center for Risk and Reliability, University of Maryland, College Park, MD 20742 (United States)

2009-05-15

This paper is a result of a research with the primary purpose of extending Probabilistic Risk Assessment (PRA) modeling frameworks to include the effects of organizational factors as the deeper, more fundamental causes of accidents and incidents. There have been significant improvements in the sophistication of quantitative methods of safety and risk assessment, but the progress on techniques most suitable for organizational safety risk frameworks has been limited. The focus of this paper is on the choice of 'representational schemes' and 'techniques.' A methodology for selecting appropriate candidate techniques and their integration in the form of a 'hybrid' approach is proposed. Then an example is given through an integration of System Dynamics (SD), Bayesian Belief Network (BBN), Event Sequence Diagram (ESD), and Fault Tree (FT) in order to demonstrate the feasibility and value of hybrid techniques. The proposed hybrid approach integrates deterministic and probabilistic modeling perspectives, and provides a flexible risk management tool for complex socio-technical systems. An application of the hybrid technique is provided in the aviation safety domain, focusing on airline maintenance systems. The example demonstrates how the hybrid method can be used to analyze the dynamic effects of organizational factors on system risk.

PRA quality and use

International Nuclear Information System (INIS)

Okrent, D.; Apostolakis, G.; Whitley, R.; Garrick, B.J.

1982-10-01

This report deals with several inter-related aspects of probabilistic risk assessment. Some prior opinion regarding quality assurance, methodology and questions of peer review are reviewed, followed by comments by the authors on these and related subjects. Problems arising in decision-making by different groups concerning the meaning and validity of a PRA are examined, and the role of performance criteria in helping to achieve consensus is treated. Finally, a general approach to the development of performance criteria for systems and functions by the retrospective comparison of existing PRAs is proposed and examined in a preliminary fashion

Use of PRA in the nuclear regulatory field in South Africa

International Nuclear Information System (INIS)

Hill, T.F.

1994-01-01

The nuclear regulatory authority in South Africa (since 1988 the Council for Nuclear Safety (CNS)), established in 1973 nuclear safety criteria against which to assess the level of safety of any facility using radioactive material. It is a regulatory requirement in South Africa to develop and maintain a living PRA for each facility and thereby to provide the necessary information to demonstrate compliance against these criteria. All safety submissions to the CNS must include at least a risk statement based on an accepted PRA study. The function of the CNS is to regulate all activities in South Africa involving the use of radioactive material and posing a significant risk to the public or plant personnel. This includes most aspects of the nuclear fuel cycle and the Koeberg NPS (two 2775 MW(th) PWRs). A PRA study including source terms for the two Koeberg units was presented by the contractor in 1979. This included the risk due to power and shutdown states and non reactor related accidents involving spent fuel storage, fuel handling and waste treatment related activities. At least 20 PRA studies have been performed for other nuclear facilities in the country. The CNS maintains an in-house PRA capability to perform independent assessments of licensee submission, to participate in developments of PRA methodology in the regulatory field, to perform pro-active safety work and to assist in regulatory decision making. Present ongoing work includes the development of a risk monitor, a risk management system, improvement in PRA codes, models, data collection and analysis, off-site risk assessment methodology and associated regulatory policy. (author). 1 fig

Fire PRA requantification studies. Final report

International Nuclear Information System (INIS)

Parkinson, W.

1993-03-01

This report describes the requantification of two existing fire probabilistic risk assessments (PRAs) using a fire PRA method and data that are being developed by the Electric Power Research Institute (EPRI). The two existing studies are the Seabrook Station Probabilistic Safety Assessment that was made in 1983 and the 1989 NUREG-1150 analysis of the Peach Bottom Plant. Except for the fire methods and data, the original assumptions were used. The results from the requantification show that there were excessive conservatisms in the original studies. The principal reason for a hundredfold reduction in the Peach Bottom core- damage frequency is the determination that no electrical cabinet fire in a switchgear room would damage both offsite power feeds. Past studies often overestimated the heat release from electrical cabinet fires. EPRI's electrical cabinet heat release rates are based on tests that were conducted for Sandia's fire research program. The rates are supported by the experience in the EPRI Fire Events Database for U.S. nuclear plants. Test data and fire event experience also removed excessive conservatisms in the Peach Bottom control and cable spreading rooms, and the Seabrook primary component cooling pump, turbine building relay and cable spreading rooms. The EPRI fire PRA method and data will show that there are excessive conservatisms in studies that were made for many plants and can benefit them accordingly

Review of seismic probabilistic risk assessment and the use of sensitivity analysis

International Nuclear Information System (INIS)

Shiu, K.K.; Reed, J.W.; McCann, M.W. Jr.

1985-01-01

This paper presents results of sensitivity reviews performed to address a range of questions which arise in the context of seismic probabilistic risk assessment (PRA). In a seismic PRA, sensitivity evaluations can be divided into three areas: hazard, fragility, and system modeling. As a part of the review of standard boiling water reactor seismic PRA which was performed by General Electric (GE), a reassessment of the plant damage states frequency and a detailed sensitivity analysis were conducted at Brookhaven National Laboratory. The rationale for such an undertaking is that in this case: (1) the standard plant may be sited anywhere in the eastern US (i.e., in regions with safety shutdown earthquake (SSE) values equal to or less than 0.3g peak ground acceleration), (2) it may have equipment whose fragility values could vary over a wide range; and (3) there are variations in system designs outside the original defined scope. Seismic event trees and fault trees were developed to model the different system and plant accident sequences. Hazard curves which represent various sites on the east coast were obtained; alternate structure and equipment fragility data were postulated. Various combinations of hazard and fragility data were analyzed. In addition, system modeling was perturbed to examine the impact upon the final results. Orders of magnitude variation were observed in the plant damage state frequency among the different cases. 7 references, 3 figures, 2 tables

Use of Probabilistic Risk Assessment in Shuttle Decision Making Process

Science.gov (United States)

Boyer, Roger L.; Hamlin, Teri, L.

2011-01-01

This slide presentation reviews the use of Probabilistic Risk Assessment (PRA) to assist in the decision making for the shuttle design and operation. Probabilistic Risk Assessment (PRA) is a comprehensive, structured, and disciplined approach to identifying and analyzing risk in complex systems and/or processes that seeks answers to three basic questions: (i.e., what can go wrong? what is the likelihood of these occurring? and what are the consequences that could result if these occur?) The purpose of the Shuttle PRA (SPRA) is to provide a useful risk management tool for the Space Shuttle Program (SSP) to identify strengths and possible weaknesses in the Shuttle design and operation. SPRA was initially developed to support upgrade decisions, but has evolved into a tool that supports Flight Readiness Reviews (FRR) and near real-time flight decisions. Examples of the use of PRA for the shuttle are reviewed.

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

International Nuclear Information System (INIS)

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

2004-01-01

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

Probabilistic risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1980-08-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the US Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed

Using level-I PRA for enhanced safety of the advanced neutron source reactor

International Nuclear Information System (INIS)

Ramsey, C.T.; Linn, M.A.

1995-01-01

The phase-1, level-I probabilistic risk assessment (PRA) of the Advanced Neutron Source (ANS) reactor has been completed as part of the conceptual design phase of this proposed research facility. Since project inception, PRA and reliability concepts have been an integral part of the design evolutions contributing to many of the safety features in the current design. The level-I PRA has been used to evaluate the internal events core damage frequency against project goals and to identify systems important to safety and availability, and it will continue to guide and provide support to accident analysis, both severe and nonsevere. The results also reflect the risk value of defense-in-depth safety features in reducing the likelihood of core damage

Probabilistic risk assessment of HTGRs

International Nuclear Information System (INIS)

Fleming, K.N.; Houghton, W.J.; Hannaman, G.W.; Joksimovic, V.

1981-01-01

Probabilistic Risk Assessment methods have been applied to gas-cooled reactors for more than a decade and to HTGRs for more than six years in the programs sponsored by the U.S. Department of Energy. Significant advancements to the development of PRA methodology in these programs are summarized as are the specific applications of the methods to HTGRs. Emphasis here is on PRA as a tool for evaluating HTGR design options. Current work and future directions are also discussed. (author)

ASSESSMENT OF DYNAMIC PRA TECHNIQUES WITH INDUSTRY AVERAGE COMPONENT PERFORMANCE DATA

Energy Technology Data Exchange (ETDEWEB)

Yadav, Vaibhav; Agarwal, Vivek; Gribok, Andrei V.; Smith, Curtis L.

2017-06-01

In the nuclear industry, risk monitors are intended to provide a point-in-time estimate of the system risk given the current plant configuration. Current risk monitors are limited in that they do not properly take into account the deteriorating states of plant equipment, which are unit-specific. Current approaches to computing risk monitors use probabilistic risk assessment (PRA) techniques, but the assessment is typically a snapshot in time. Living PRA models attempt to address limitations of traditional PRA models in a limited sense by including temporary changes in plant and system configurations. However, information on plant component health are not considered. This often leaves risk monitors using living PRA models incapable of conducting evaluations with dynamic degradation scenarios evolving over time. There is a need to develop enabling approaches to solidify risk monitors to provide time and condition-dependent risk by integrating traditional PRA models with condition monitoring and prognostic techniques. This paper presents estimation of system risk evolution over time by integrating plant risk monitoring data with dynamic PRA methods incorporating aging and degradation. Several online, non-destructive approaches have been developed for diagnosing plant component conditions in nuclear industry, i.e., condition indication index, using vibration analysis, current signatures, and operational history [1]. In this work the component performance measures at U.S. commercial nuclear power plants (NPP) [2] are incorporated within the various dynamic PRA methodologies [3] to provide better estimates of probability of failures. Aging and degradation is modeled within the Level-1 PRA framework and is applied to several failure modes of pumps and can be extended to a range of components, viz. valves, generators, batteries, and pipes.

Calculation of Fire Severity Factors and Fire Non-Suppression Probabilities For A DOE Facility Fire PRA

International Nuclear Information System (INIS)

Elicson, Tom; Harwood, Bentley; Lucek, Heather; Bouchard, Jim

2011-01-01

Over a 12 month period, a fire PRA was developed for a DOE facility using the NUREG/CR-6850 EPRI/NRC fire PRA methodology. The fire PRA modeling included calculation of fire severity factors (SFs) and fire non-suppression probabilities (PNS) for each safe shutdown (SSD) component considered in the fire PRA model. The SFs were developed by performing detailed fire modeling through a combination of CFAST fire zone model calculations and Latin Hypercube Sampling (LHS). Component damage times and automatic fire suppression system actuation times calculated in the CFAST LHS analyses were then input to a time-dependent model of fire non-suppression probability. The fire non-suppression probability model is based on the modeling approach outlined in NUREG/CR-6850 and is supplemented with plant specific data. This paper presents the methodology used in the DOE facility fire PRA for modeling fire-induced SSD component failures and includes discussions of modeling techniques for: Development of time-dependent fire heat release rate profiles (required as input to CFAST), Calculation of fire severity factors based on CFAST detailed fire modeling, and Calculation of fire non-suppression probabilities.

The accidents during shutdown conditions Temelin NPP

International Nuclear Information System (INIS)

Sykora, M.; Mlady, O.

1996-01-01

Two parallel activities oriented for the accidents during shutdown conditions are performed at Temelin NPP: Development of symptom based emergency operating procedures (EOPs) applicable for the accidents which could occur during operational modes 1 through 4; independent evaluation of plant safety as part of the Temelin Shutdown probabilistic assessment to define the accidents which could occur during mode 5 and 6 for which the EOPs must be extended. Both these activities are in progress now because Temelin plant is still in the construction phase

PRA: an evaluation of state-of-the-art

International Nuclear Information System (INIS)

Joksimovich, V.

1985-01-01

Some elements of the probabilistic risk assessment (PRA) methodology can be characterized as mature and are even ready for some kind of a standardization effort. Other elements are still, however, in a rapid state of evolution. Questions are continuously being asked regarding maturity of PRA techniques vis-a-vis a regulatory decision-making process. Establishing a framework for evaluating state-of-the-art in any technological field is a challenging task. An implementation of a selected framework to a satisfactory conclusion is a monumental task. Of course, these types of issues can be discussed meaningfully only if they are tied to a particular application. The author's participation in the NSF-sponsored risk assessment project is discussed in the paper. The evaluation employed here makes use of the following five evaluation criteria: logical soundness, completeness, accuracy, acceptability, and practicality

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

Innovative probabilistic risk assessment applications: barrier impairments and fracture toughness. Panel Discussion

International Nuclear Information System (INIS)

Osterman, Michael; Root, Steven; Li, F.; Modarres, Mohammad; Reinhart, F. Mark; Bradley, Biff; Calhoun, David J.

2001-01-01

Full text of publication follows: New probabilistic risk assessment (PRA) applications promise to improve the overall safety and efficiency of nuclear plant operations. This discussion will explore the use of PRA in evaluating barrier integrity with respect to the consequences of natural phenomena such as tornadoes, floods, and harsh environments. Additionally, the session will explore proposals to improve fracture toughness techniques using PRA. (authors)

ISSUES ASSOCIATED WITH PROBABILISTIC FAILURE MODELING OF DIGITAL SYSTEMS

International Nuclear Information System (INIS)

CHU, T.L.; MARTINEZ-GURIDI, G.; LIHNER, J.; OVERLAND, D.

2004-01-01

The current U.S. Nuclear Regulatory Commission (NRC) licensing process of instrumentation and control (I and C) systems is based on deterministic requirements, e.g., single failure criteria, and defense in depth and diversity. Probabilistic considerations can be used as supplements to the deterministic process. The National Research Council has recommended development of methods for estimating failure probabilities of digital systems, including commercial off-the-shelf (COTS) equipment, for use in probabilistic risk assessment (PRA). NRC staff has developed informal qualitative and quantitative requirements for PRA modeling of digital systems. Brookhaven National Laboratory (BNL) has performed a review of the-state-of-the-art of the methods and tools that can potentially be used to model digital systems. The objectives of this paper are to summarize the review, discuss the issues associated with probabilistic modeling of digital systems, and identify potential areas of research that would enhance the state of the art toward a satisfactory modeling method that could be integrated with a typical probabilistic risk assessment

Applications of the EBR-II Probabilistic Risk Assessment

International Nuclear Information System (INIS)

Roglans, J.: Ragland, W.A.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future

A methodology for reviewing Probabilistic Risk Assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase. The value of this procedure comes from having a systematic framework for the PRA review. Practical constraints limit the time and qualified people needed for an adequate review. Having the established framework from this procedure as a starting point, reviewers can focus most of their attention on the accuracy and the completeness of the calculation. Time wasted at the start of the review is reduced by first using this procedure to sort through the technical details of the PRA and to reconstruct the risk estimate from dominant contributions

PRA research and the development of risk-informed regulation at the U.S. nuclear regulatory commission

International Nuclear Information System (INIS)

Siu, Nathan; Collins, Dorothy

2008-01-01

Over the years, Probabilistic Risk Assessment (PRA) research activities conducted at the U.S. Nuclear Regulatory Commission (NRC) have played an essential role in support of the agency's move towards risk-informed regulation. These research activities have provided the technical basis for NRC's regulatory activities in key areas; provided PRA methods, tools, and data enabling the agency to meet future challenges; supported the implementation of NRC's 1995 PRA Policy Statement by assessing key sources of risk; and supported the development of necessary technical and human resources supporting NRC's risk-informed activities. PRA research aimed at improving the NRC's understanding of risk can positively affect the agency's regulatory activities, as evidenced by three case studies involving research on fire PRA, Human Reliability Analysis (HRA), and Pressurized Thermal Shock (PTS) PRA. These case studies also show that such research can take a considerable amount of time, and that the incorporation of research results into regulatory practice can take even longer. The need for sustained effort and appropriate lead time is an important consideration in the development of a PRA research program aimed at helping the agency address key sources of risk for current and potential future facilities

Probabilistic risk assessment: A look at the role of artificial intelligence

International Nuclear Information System (INIS)

Wang, J.; Modarres, M.; Hunt, R.N.M.

1988-01-01

A review of traditional Probabilistic Risk Assessment (PRA) methods used in the nuclear power industry is presented. The shortcomings of the current PRA methods are pointed out. A method of performing a PRA is proposed and is computerized. The role of artificial intelligence in developing and performing the proposed PRA approach is discussed. The proposed PRA approach is verified by comparing the results to previously performed PRAs. The comparisons have supported the adequacy and completeness of the results of the proposed model. A discussion of how the proposed method can be used as an expert system to verify plant status following loss of plant hardware is also presented. (orig.)

A methodology for reviewing probabilistic risk assessments

International Nuclear Information System (INIS)

Derby, S.L.

1983-01-01

The starting point for peer review of a Probabilistic Risk Assessment (PRA) is a clear understanding of how the risk estimate was prepared and of what contributions dominate the calculation. The problem facing the reviewers is how to cut through the complex details of a PRA to gain this understanding. This paper presents a structured, analytical procedure that solves this problem. The effectiveness of this solution is demonstrated by an application on the Zion Probabilistic Safety Study. The procedure found the three dominant initiating events and provided a simplified reconstruction of the calculation of the risk estimate. Significant assessments of uncertainty were also identified. If peer review disputes the accuracy of these judgments, then the revised risk estimate could significantly increase

PRA -- Now that operators have it, what do they do with it?

International Nuclear Information System (INIS)

Rasmussen, M.A.; Kolo, R.J.

1996-01-01

Many utilities have had Probabilistic Risk Assessment (PRA) projects underway for several years in order to satisfy the NRC Generic Letter 88-20 requirement for an Individual Plant Examination, or IPE. Typically the studies have reached the conclusion that there are significant differences in the contribution of different plant components to preventing core damage should a major plant transient occur. How nuclear plant operators can use this knowledge to DECREASE the overall risk of performing the routine tasks of testing and maintenance is not an easy task. 10CFR50.65; ''The Maintenance Rule,'' requires that any plant maintenance performed with the unit on line be evaluated for risk. Byron Station will satisfy the 10CFR50.65 requirement by using PRA methodology to evaluate testing and maintenance activities performed with the unit at power. The challenge is to effectively use the results of PRA studies to aid in plant operations without having to make on shift plant operations personnel experts in PRA. At Byron, PRA is used to help build the weekly work schedules. Operations personnel tasked with reviewing the work schedule are the departmental experts on the use of the PRA results. The on shift SRO's role in implementing the program is to accurately execute and monitor the work week schedule as written, and to react to unforeseen equipment failures with an appropriate level of response. The response to such emergent work items is also predefined. Handling emergent work in a prescribed manner minimizes the overall risk to the unit and also eliminates the need to have PRA expertise available to make emergent work risk evaluations. Thus the on shift operators' required knowledge of PRA methods and intricacies is minimized. PRA is just another of the many tools used by the shift operator to run the plant in a safe, conservative manner

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

Energy Technology Data Exchange (ETDEWEB)

Monteleone, S. [comp.] [Brookhaven National Lab., Upton, NY (United States)

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database.

Proceedings of the twenty-fourth water reactor safety information meeting. Volume 3: PRA and HRA; Probabilistic seismic hazard assessment and seismic siting criteria

International Nuclear Information System (INIS)

Monteleone, S.

1997-02-01

This three-volume report contains papers presented at the Twenty-Fourth Water Reactor Safety Information Meeting held at the Bethesda Marriott Hotel, Bethesda, Maryland, October 21--23, 1996. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included papers presented by researchers from Czech Republic, Finland, France, Japan, Norway, Russia and United Kingdom. This volume is divided into the following sections: PRA and HRA and probabilistic seismic hazard assessment and seismic siting criteria. Selected papers are indexed separately for inclusion in the Energy Science and Technology Database

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

Energy Technology Data Exchange (ETDEWEB)

Kang, Daeil; Kim, J. H.; Jang, S. C

2007-03-15

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post

A Human Reliability Analysis of Post- Accident Human Errors in the Low Power and Shutdown PSA of KSNP

International Nuclear Information System (INIS)

Kang, Daeil; Kim, J. H.; Jang, S. C.

2007-03-01

Korea Atomic Energy Research Institute, using the ANS low power and shutdown (LPSD) probabilistic risk assessment (PRA) Standard, evaluated the LPSD PSA model of the KSNP, Yonggwang Units 5 and 6, and identified the items to be improved. The evaluation results of human reliability analysis (HRA) of the post-accident human errors in the LPSD PSA model for the KSNP showed that 10 items among 19 items of supporting requirements for those in the ANS PRA Standard were identified as them to be improved. Thus, we newly carried out a HRA for post-accident human errors in the LPSD PSA model for the KSNP. Following tasks are the improvements in the HRA of post-accident human errors of the LPSD PSA model for the KSNP compared with the previous one: Interviews with operators in the interpretation of the procedure, modeling of operator actions, and the quantification results of human errors, site visit. Applications of limiting value to the combined post-accident human errors. Documentation of information of all the input and bases for the detailed quantifications and the dependency analysis using the quantification sheets The assessment results for the new HRA results of post-accident human errors using the ANS LPSD PRA Standard show that above 80% items of its supporting requirements for post-accident human errors were graded as its Category II. The number of the re-estimated human errors using the LPSD Korea Standard HRA method is 385. Among them, the number of individual post-accident human errors is 253. The number of dependent post-accident human errors is 135. The quantification results of the LPSD PSA model for the KSNP with new HEPs show that core damage frequency (CDF) is increased by 5.1% compared with the previous baseline CDF It is expected that this study results will be greatly helpful to improve the PSA quality for the domestic nuclear power plants because they have sufficient PSA quality to meet the Category II of Supporting Requirements for the post

Application of probabilistic risk assessment in the operation of Koeberg nuclear power station

International Nuclear Information System (INIS)

Nicholls, D.R.

1991-01-01

Probabilistic risk assessment (PRA) calculates the probability that a set of multiple failures could occur, the frequency with which the safety circuits will be required and the consequences of the failure of the safety systems. In this way the frequency with which major accident situations can be expected to happen, can be derived. The world history of PRA is presented, together with the South African history of PRA. The theory of PRA is explained and the application of PRA studies is described. In the last twenty years, PRA has gone from being a theoretical idea to a practical tool for assisting in plant management. 2 figs., 1 ill

Use of plant-specific PRA in an EOP scope audit

International Nuclear Information System (INIS)

O'Brien, J.J.

1991-01-01

Traditionally, decisions on which accident scenarios to proceduralize as emergency operating procedures (EOPs) have been based on existing design basis analyses, engineering judgment, and probabilistic risk assessments (PRAs) on generic plants. This approach has important strengths and limits. The major limitation of generic PRAs is their inability to account for plant-specific features. Use of plant-specific PRA to determine the impact of proceduralizing, or not proceduralizing, responses to scenarios considers plant-specific features. This helps to eliminate unnecessary EOPs, thus allowing resources to be concentrated on scenarios that are more important for a particular plant. In preparation for a US Nuclear Regulatory Commission audit, a plant-specific PRA was used to assess and quantify the plant's previous decision not to implement six reference emergency response guidelines (ERGs) as procedures. The original justification for nonimplementation of the ERGs was based on engineering judgment. The PRA provided a quantitative justification for implementation/nonimplementation of each guidelines. This analysis accounted for plant-specific design features not common to all reference plants

Task analysis: How far are we from usable PRA input

International Nuclear Information System (INIS)

Gertman, D.I.; Blackman, H.S.; Hinton, M.F.

1984-01-01

This chapter reviews data collected at the Idaho National Engineering Laboratory for three DOE-owned reactors (the Advanced Test Reactor, the Power Burst Facility, and the Loss of Fluids Test Reactor) in order to identify usable Probabilistic Risk Assessment (PRA) input. Task analytic procedures involve the determination of manning and skill levels as a means of determining communication requirements, in assessing job performance aids, and in assessing the accuracy and completeness of emergency and maintenance procedures. The least understood aspect in PRA and plant reliability models is the human factor. A number of examples from the data base are discussed and offered as a means of providing more meaningful data than has been available to PRA analysts in the past. It is concluded that the plant hardware-procedures-personnel interfaces are essential to safe and efficient plant operations and that task analysis is a reasonably sound way of achieving a qualitative method for identifying those tasks most strongly associated with task difficulty, severity of consequence, and error probability

Bayesian parameter estimation in probabilistic risk assessment

International Nuclear Information System (INIS)

Siu, Nathan O.; Kelly, Dana L.

1998-01-01

Bayesian statistical methods are widely used in probabilistic risk assessment (PRA) because of their ability to provide useful estimates of model parameters when data are sparse and because the subjective probability framework, from which these methods are derived, is a natural framework to address the decision problems motivating PRA. This paper presents a tutorial on Bayesian parameter estimation especially relevant to PRA. It summarizes the philosophy behind these methods, approaches for constructing likelihood functions and prior distributions, some simple but realistic examples, and a variety of cautions and lessons regarding practical applications. References are also provided for more in-depth coverage of various topics

Method and system for dynamic probabilistic risk assessment

Science.gov (United States)

Dugan, Joanne Bechta (Inventor); Xu, Hong (Inventor)

2013-01-01

The DEFT methodology, system and computer readable medium extends the applicability of the PRA (Probabilistic Risk Assessment) methodology to computer-based systems, by allowing DFT (Dynamic Fault Tree) nodes as pivot nodes in the Event Tree (ET) model. DEFT includes a mathematical model and solution algorithm, supports all common PRA analysis functions and cutsets. Additional capabilities enabled by the DFT include modularization, phased mission analysis, sequence dependencies, and imperfect coverage.

Organizational extension of PRA models and NASA application

International Nuclear Information System (INIS)

Pate-Cornell, E.

1989-01-01

This paper describes a probabilistic method which extends classical PRA to include some characteristics of the organization that processes or manages an engineering system. Ataxonomy of errors is presented and their organizational roots are examined. An assembly model is proposed for the analysis of the resulting spectrum of capacities of the system. The management of the Thermal Protection system of the Space Shuttle is used as an illustration. The model allows assessment of the benefits of organizational improvements of the orbiter's processing

Overview of seismic probabilistic risk assessment for structural analysis in nuclear facilities

International Nuclear Information System (INIS)

Reed, J.W.

1989-01-01

Probabilistic Risk Assessment (PRA) for seismic events is currently being performed for nuclear and DOE facilities. The background on seismic PRA is presented along with a basic description of the method. The seismic PRA technique is applicable to other critical facilities besides nuclear plants. The different approaches for obtained structure fragility curves are discussed and their applications to structures and equipment, in general, are addressed. It is concluded that seismic PRA is a useful technique for conducting probability analysis for a wide range of classes of structures and equipment

Risk impact of BWR technical specifications requirements during shutdown

International Nuclear Information System (INIS)

Staple, B.D.; Kirk, H.K.; Yakle, J.

1994-10-01

This report presents an application of probabilistic models and risk based criteria for determining the risk impact of the Limiting Conditions of Operations (LCOs) in the Technical Specifications (TSs) of a boiling water reactor during shutdown. This analysis studied the risk impact of the current requirements of Allowed Outage Times (AOTs) and Surveillance Test Intervals (STIs) in eight Plant Operational States (POSs) which encompass power operations, shutdown, and refueling. This report also discusses insights concerning TS action statements

The Evaluation of the Adequacy of PRA Results for Risk-informed Decision Makings With Respect to Incompleteness

International Nuclear Information System (INIS)

Kang, Kyungmin; Jae, Moosung

2007-01-01

PRA(Probabilistic Risk Assessment), as a quantitative tool, has many strengths as well as weaknesses. There are several limitations on the use of PRA techniques for risk modeling and analysis. First, the true values of most model inputs are unknown. Ideally, probability distribution models are well developed and assigned to the unknown input parameters to reflect the analyst's state of knowledge of the values of this input parameter. The problem of overconfidence and lack of confidence in the values of certain model input parameters can lead to inaccurate PRA results. Secondly, the analyst's lack of knowledge of a system's practical application as opposed to its theoretical operation can lead to modeling errors. The quality of PRAs has been addressed by a number of regulatory and industry organizations Some have argued that a good PRA should be a complete, full scope, three level PRA, while others have claimed that the quality of a PRA should be measured with respect to the application and decision supported. we show by way of an example that the adequacy of a PRA results is important to risk-informed decision making process and should be measured with respect to the application and decision supported

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs

Use of probabilistic risk assessment in fuel cycle facilities

International Nuclear Information System (INIS)

Gonzalez, Felix; Gonzalez, Michelle; Wagner, Brian

2013-01-01

As expressed in its Policy Statement on the Use of Probabilistic Risk Assessment (PRA) Methods in Nuclear Regulatory Activities, the U.S Nuclear Regulatory Commission has been working for decades to increase the use of PRA technology in its regulatory activities. Since the policy statement was issued in 1995, PRA has become a core component of the nuclear power plant (NPP) licensing and oversight processes. In the last several years, interest has increased in PRA technologies and their possible application to other areas including, but not limited to, spent fuel handling, fuel cycle facilities, reprocessing facilities, and advanced reactors. This paper describes the application of PRA technology currently used in NPPs and its application in other areas such as fuel cycle facilities and advanced reactors. It describes major challenges that are being faced in the application of PRA into new technical areas and possible ways to resolve them. (authors)

Augmenting Probabilistic Risk Assesment with Malevolent Initiators

International Nuclear Information System (INIS)

Smith, Curtis; Schwieder, David

2011-01-01

As commonly practiced, the use of probabilistic risk assessment (PRA) in nuclear power plants only considers accident initiators such as natural hazards, equipment failures, and human error. Malevolent initiators are ignored in PRA, but are considered the domain of physical security, which uses vulnerability assessment based on an officially specified threat (design basis threat). This paper explores the implications of augmenting and extending existing PRA models by considering new and modified scenarios resulting from malevolent initiators. Teaming the augmented PRA models with conventional vulnerability assessments can cost-effectively enhance security of a nuclear power plant. This methodology is useful for operating plants, as well as in the design of new plants. For the methodology, we have proposed an approach that builds on and extends the practice of PRA for nuclear power plants for security-related issues. Rather than only considering 'random' failures, we demonstrated a framework that is able to represent and model malevolent initiating events and associated plant impacts.

Plant Operation Station for HTR-PM Low Power and Shutdown operation Probabilistic safety analysis

International Nuclear Information System (INIS)

Liu Tao; Tong Jiejuan

2014-01-01

Full range Probabilistic safety analysis (PSA) is one of key conditions for nuclear power plant (NPP) licensing according to the requirement of nuclear safety regulatory authority. High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) has developed construction design and prepared for the charging license application. So after the normal power operation PSA submitted for review, the Low power and Shutdown operation Probabilistic safety analysis (LSPSA) also begin. The results of LSPSA will together with prior normal power PSA results to demonstrate the safety level of HTR-PM NPP Plant Operation Station (POS) is one of important terms in LSPSA. The definition of POS lays the foundation for LSPSA modeling. POS provides initial and boundary conditions for the following event tree and fault tree model development. The aim of this paper is to describe the state-of-the-art of POS definition for HTR-PM LSPSA. As for the first attempt to the high temperature gas cooled reactor module plant, the methodology and procedure of POS definition refers to the LWR LSPSA guidance, and adds to plant initial status analysis due to the HTR-PM characteristics. A specific set of POS grouping vectors is investigate and suggested for HTR-PM NPP, which reflects the characteristics of plant modularization and on-line refueling. As a result, seven POSs are given according to the grouping vectors at the end of the paper. They will be used to the LSPSA modelling and adjusted if necessary. The papers ’work may provide reference to the analogous NPP LSPSA. (author)

A model for assessing human cognitive reliability in PRA studies

International Nuclear Information System (INIS)

Hannaman, G.W.; Spurgin, A.J.; Lukic, Y.

1985-01-01

This paper summarizes the status of a research project sponsored by EPRI as part of the Probabilistic Risk Assessment (PRA) technology improvement program and conducted by NUS Corporation to develop a model of Human Cognitive Reliability (HCR). The model was synthesized from features identified in a review of existing models. The model development was based on the hypothesis that the key factors affecting crew response times are separable. The inputs to the model consist of key parameters the values of which can be determined by PRA analysts for each accident situation being assessed. The output is a set of curves which represent the probability of control room crew non-response as a function of time for different conditions affecting their performance. The non-response probability is then a contributor to the overall non-success of operating crews to achieve a functional objective identified in the PRA study. Simulator data and some small scale tests were utilized to illustrate the calibration of interim HCR model coefficients for different types of cognitive processing since the data were sparse. The model can potentially help PRA analysts make human reliability assessments more explicit. The model incorporates concepts from psychological models of human cognitive behavior, information from current collections of human reliability data sources and crew response time data from simulator training exercises

A review of NRC staff uses of probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC`s Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff`s current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff`s uses of PRA.

A review of NRC staff uses of probabilistic risk assessment

International Nuclear Information System (INIS)

1994-03-01

The NRC staff uses probabilistic risk assessment (PRA) and risk management as important elements its licensing and regulatory processes. In October 1991, the NRC's Executive Director for Operations established the PRA Working Group to address concerns identified by the Advisory Committee on Reactor Safeguards with respect to unevenness and inconsistency in the staff's current uses of PRA. After surveying current staff uses of PRA and identifying needed improvements, the Working Group defined a set of basic principles for staff PRA use and identified three areas for improvements: guidance development, training enhancements, and PRA methods development. For each area of improvement, the Working Group took certain actions and recommended additional work. The Working Group recommended integrating its work with other recent PRA-related activities the staff completed and improving staff interactions with PRA users in the nuclear industry. The Working Group took two key actions by developing general guidance for two uses of PRA within the NRC (that is, screening or prioritizing reactor safety issues and analyzing such issues in detail) and developing guidance on basic terms and methods important to the staff's uses of PRA

Seismic PRA of a BWR plant

International Nuclear Information System (INIS)

Nishio, Masahide; Fujimoto, Haruo

2014-01-01

Since the occurrence of nuclear power plant accidents in the Fukushima Daichi nuclear power station, the regulatory framework on severe accident (SA) has been discussed in Japan. The basic concept is to typify and identify the accident sequences leading to core/primary containment vessel (PCV) damage and to implement SA measures covering internal and external events extensively. As Japan is an earthquake-prone country and earthquakes and tsunami are important natural external events for nuclear safety of nuclear power plants, JNES performed the seismic probabilistic risk assessment (PRA) on a typical nuclear power plant and evaluated the dominant accident sequences leading to core/PCV damage to discuss dominant scenarios of severe accident (SA). The analytical models and the results of level-1 seismic PRA on a 1,100 MWe BWR-5 plant are shown here. Seismic PRA was performed for a typical BWR5 plant. Initiating events with large contribution to core damage frequency are the loss of all AC powers (station blackout) and the large LOCA. The top of dominant accident sequences is the simultaneous occurrence of station blackout and large LOCA. Important components to core damage frequency are electric power supply equipment. It needs to keep in mind that the results are influenced on site geologic characteristic to a greater or lesser. In the process of analysis, issues such as conservative assumptions related to damages of building or structure and success criteria for excessive LOCA are left to be resolved. These issues will be further studied including thermal hydric analysis in the future. (authors)

A perspective of PC-based probabilistic risk assessment

International Nuclear Information System (INIS)

Sattison, M.B.; Rasmuson, D.M.; Robinson, R.C.; Russell, K.D.; Van Siclen, V.S.

1987-01-01

Probabilistic risk assessment (PRA) information has been under-utilized in the past due to the large effort required to input the PRA data and the large expense of the computers needed to run PRA codes. The microcomputer-based Integrated Reliability and Risk Analysis System (IRRAS) and the System Analysis and Risk Assessment (SARA) System, under development at the Idaho National Engineering Laboratory, have greatly enhanced the ability of managers to use PRA techniques in their decision-making. IRRAS is a tool that allows an analyst to create, modify, update, and reanalyze a plant PRA to keep the risk assessment current with the plant's configuration and operation. The SARA system is used to perform sensitivity studies on the results of a PRA. This type of analysis can be used to evaluate proposed changes to a plant or its operation. The success of these two software projects demonstrate that risk information can be made readily available to those that need it. This is the first step in the development of a true risk management capability

Application of the NUREG/CR-6850 EPRI/NRC Fire PRA Methodology to a DOE Facility

International Nuclear Information System (INIS)

Elicson, Tom; Harwood, Bentley; Yorg, Richard; Lucek, Heather; Bouchard, Jim; Jukkola, Ray; Phan, Duan

2011-01-01

The application NUREG/CR-6850 EPRI/NRC fire PRA methodology to DOE facility presented several challenges. This paper documents the process and discusses several insights gained during development of the fire PRA. A brief review of the tasks performed is provided with particular focus on the following: Tasks 5 and 14: Fire-induced risk model and fire risk quantification. A key lesson learned was to begin model development and quantification as early as possible in the project using screening values and simplified modeling if necessary. Tasks 3 and 9: Fire PRA cable selection and detailed circuit failure analysis. In retrospect, it would have been beneficial to perform the model development and quantification in 2 phases with detailed circuit analysis applied during phase 2. This would have allowed for development of a robust model and quantification earlier in the project and would have provided insights into where to focus the detailed circuit analysis efforts. Tasks 8 and 11: Scoping fire modeling and detailed fire modeling. More focus should be placed on detailed fire modeling and less focus on scoping fire modeling. This was the approach taken for the fire PRA. Task 14: Fire risk quantification. Typically, multiple safe shutdown (SSD) components fail during a given fire scenario. Therefore dependent failure analysis is critical to obtaining a meaningful fire risk quantification. Dependent failure analysis for the fire PRA presented several challenges which will be discussed in the full paper.

Risk management on nuclear power plant. Application of probabilistic risk assessment

International Nuclear Information System (INIS)

Kojima, Shigeo

2003-01-01

In U.S.A., nuclear safety regulation is moving to risk-informed regulation (RIR), so necessity of a standard to provide contents of probabilistic risk assessment (PRA) constructing its roots has been discussed for a long time. In 1998, the Committee on Nuclear Risk Management (CNRM) of the American Society of Mechanical Engineers (ASME) began to investigate the standard, of which last edition was published as the Standard for Probabilistic Risk Management for Nuclear Power Plant Applications: RA-S-2002 (PRMA) on April, 2002. As in the Committee, the Nuclear Regulatory Commission (NRC), electric power companies, national institutes, PRA specialists, and so on took parts to carry out many discussions with full energies of participants on risk management in U.S.A., the standard was finished after about four years' efforts. In U.S.A., risk management having already used PRA is successfully practiced, U.S.A. is at a stage with more advancing steps of the risk management than Japan is. Here was described on the standard of PRA and a concrete method of the risk management carried out at nuclear power stations. (G.K.)

PRA and Conceptual Design

Science.gov (United States)

DeMott, Diana; Fuqua, Bryan; Wilson, Paul

2013-01-01

Once a project obtains approval, decision makers have to consider a variety of alternative paths for completing the project and meeting the project objectives. How decisions are made involves a variety of elements including: cost, experience, current technology, ideologies, politics, future needs and desires, capabilities, manpower, timing, available information, and for many ventures management needs to assess the elements of risk versus reward. The use of high level Probabilistic Risk Assessment (PRA) Models during conceptual design phases provides management with additional information during the decision making process regarding the risk potential for proposed operations and design prototypes. The methodology can be used as a tool to: 1) allow trade studies to compare alternatives based on risk, 2) determine which elements (equipment, process or operational parameters) drives the risk, and 3) provide information to mitigate or eliminate risks early in the conceptual design to lower costs. Creating system models using conceptual design proposals and generic key systems based on what is known today can provide an understanding of the magnitudes of proposed systems and operational risks and facilitates trade study comparisons early in the decision making process. Identifying the "best" way to achieve the desired results is difficult, and generally occurs based on limited information. PRA provides a tool for decision makers to explore how some decisions will affect risk before the project is committed to that path, which can ultimately save time and money.

Review process and quality assurance in the EBR-II probabilistic risk assessment

International Nuclear Information System (INIS)

Roglans, J.; Hill, D.J.; Ragland, W.A.

1992-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A reactor, has recently been completed at Argonne National Laboratory (ANL). Within the scope of the ANL QA Programs, a QA Plan specifically for the EBR-II PRA was developed. The QA Plan covered all aspects of the PRA development, with emphasis on the procedures for document and software control, and the internal and external review process. The effort spent in the quality assurance tasks for the EBR-II PRA has reciprocated by providing acceptance of the work and confidence in the quality of the results

Uncertainty analysis in the applications of nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Le Duy, T.D.

2011-01-01

The aim of this thesis is to propose an approach to model parameter and model uncertainties affecting the results of risk indicators used in the applications of nuclear Probabilistic Risk assessment (PRA). After studying the limitations of the traditional probabilistic approach to represent uncertainty in PRA model, a new approach based on the Dempster-Shafer theory has been proposed. The uncertainty analysis process of the proposed approach consists in five main steps. The first step aims to model input parameter uncertainties by belief and plausibility functions according to the data PRA model. The second step involves the propagation of parameter uncertainties through the risk model to lay out the uncertainties associated with output risk indicators. The model uncertainty is then taken into account in the third step by considering possible alternative risk models. The fourth step is intended firstly to provide decision makers with information needed for decision making under uncertainty (parametric and model) and secondly to identify the input parameters that have significant uncertainty contributions on the result. The final step allows the process to be continued in loop by studying the updating of beliefs functions given new data. The proposed methodology was implemented on a real but simplified application of PRA model. (author)

Probabilistic commentary: the rise and fall, and rise again, of risk assessment

International Nuclear Information System (INIS)

Hendrie, J.M.

1985-02-01

Probabilistic risk assessment is mainly concerned with assessing the risks of nuclear power plants. Historically, the field of PRA began with a Senate request for a report on the safety of nuclear reactors in 1972. A quantitative report called WASH-1400 was eventually prepared and published in 1975, and in summary, it stated that nuclear reactors warranted only a low-grade concern in modern society. Criticism of this report and public perception of its results were highly visible subjects in the media, and the criticism led to the fact that PRA fell into disfavor. After Three Mile Island, it was recognized that PRA was a valuable tool for understanding such accidents, and PRA became a bit more popular again by the end of 1979. The usefulness of PRA was also supported by a German study in 1979. PRA played a significant role in the hearings on the Indian Point reactor. The present NRC regards PRA as an important tool in regulatory practice

Cognitive modeling and dynamic probabilistic simulation of operating crew response to complex system accidents

International Nuclear Information System (INIS)

Chang, Y.H.J.; Mosleh, A.

2007-01-01

This is the last in a series of five papers that discuss the Information Decision and Action in Crew (IDAC) context for human reliability analysis (HRA) and example application. The model is developed to probabilistically predict the responses of the control room operating crew in nuclear power plants during an accident, for use in probabilistic risk assessments (PRA). The operator response spectrum includes cognitive, emotional, and physical activities during the course of an accident. This paper describes a dynamic PRA computer simulation program, accident dynamics simulator (ADS), developed in part to implement the IDAC model. This paper also provides a detailed example of implementing a simpler version of IDAC, compared with the IDAC model discussed in the first four papers of this series, to demonstrate the practicality of integrating a detailed cognitive HRA model within a dynamic PRA framework

SEPRA - shutdown PSA for the OLKILUOTO nuclear power plant

International Nuclear Information System (INIS)

Himanen, R.

1995-01-01

The utility TVO has extended the PSA study to the analysis of refueling, shutdown and startup. The Shutdown Event PRA (SEPRA) was reported to the authority in September 1992. The study consists of the analysis of leaks and loss of decay heat removal in the planned shutdown conditions. Special studies were performed for the cold pressurization, for local criticality events, for heavy load transport and for the transients during startup and shutdown. A remarkable effort was put to identify risks, i.e. to the qualitative analysis. The regular preventive maintenance tasks in the refueling outages were analyzed and the important tasks were selected for further studies. Besides the severe core damage risk the utility was interested in less grave consequences, e.g. the economic risks, causing significant extension of outages. The plant specific screening of initiators consisted of a study on the incident history and of interviewing the plant personnel on selected tasks. A number of thermohydraulic calculations were carried out to support the analysis of accident sequences. The operator actions after an initiating event were verified with the operating staff. The annual core damage risk from the refueling outage is about one forth of the total annual risk. The modifications decreased significantly the core damage frequency. It is foreseen that the SEPRA will form a basis of the procedure enhancement for the low power states. (author) 5 figs., 1 tab., 10 refs

Computer aided probabilistic assessment of containment integrity

International Nuclear Information System (INIS)

Tsai, J.C.; Touchton, R.A.

1984-01-01

In the probabilistic risk assessment (PRA) of a nuclear power plant, there are three probability-based techniques which are widely used for event sequence frequency quantification (including nodal probability estimation). These three techniques are the event tree analysis, the fault tree analysis and the Bayesian approach for database development. In the barrier analysis for assessing radionuclide release to the environment in a PRA study, these techniques are employed to a greater extent in estimating conditions which could lead to failure of the fuel cladding and the reactor coolant system (RCS) pressure boundary, but to a lesser degree in the containment pressure boundary failure analysis. The main reason is that containment issues are currently still in a state of flux. In this paper, the authors describe briefly the computer programs currently used by the nuclear industry to do event tree analyses, fault tree analyses and the Bayesian update. The authors discuss how these computer aided probabilistic techniques might be adopted for failure analysis of the containment pressure boundary

A framework to integrate software behavior into dynamic probabilistic risk assessment

International Nuclear Information System (INIS)

Zhu Dongfeng; Mosleh, Ali; Smidts, Carol

2007-01-01

Software plays an increasingly important role in modern safety-critical systems. Although, research has been done to integrate software into the classical probabilistic risk assessment (PRA) framework, current PRA practice overwhelmingly neglects the contribution of software to system risk. Dynamic probabilistic risk assessment (DPRA) is considered to be the next generation of PRA techniques. DPRA is a set of methods and techniques in which simulation models that represent the behavior of the elements of a system are exercised in order to identify risks and vulnerabilities of the system. The fact remains, however, that modeling software for use in the DPRA framework is also quite complex and very little has been done to address the question directly and comprehensively. This paper develops a methodology to integrate software contributions in the DPRA environment. The framework includes a software representation, and an approach to incorporate the software representation into the DPRA environment SimPRA. The software representation is based on multi-level objects and the paper also proposes a framework to simulate the multi-level objects in the simulation-based DPRA environment. This is a new methodology to address the state explosion problem in the DPRA environment. This study is the first systematic effort to integrate software risk contributions into DPRA environments

Probabilistic Analysis of Failures Mechanisms of Large Dams

NARCIS (Netherlands)

Shams Ghahfarokhi, G.

2014-01-01

Risk and reliability analysis is presently being performed in almost all fields of engineering depending upon the specific field and its particular area. Probabilistic risk analysis (PRA), also called quantitative risk analysis (QRA) is a central feature of hydraulic engineering structural design.

On the functional failures concept and probabilistic safety margins: challenges in application for evaluation of effectiveness of shutdown systems - 15318

International Nuclear Information System (INIS)

Serghiuta, D.; Tholammakkil, J.

2015-01-01

The use of level-3 reliability approach and the concept of functional failure probability could provide the basis for defining a safety margin metric which would include a limit for the probability of functional failure, in line with the definition of a reliability-based design. It can also allow a quantification of level of confidence, by explicit modeling and quantification of uncertainties, and provide a better framework for representation of actual design and optimization of design margins within an integrated probabilistic-deterministic model. This paper reviews the attributes and challenges in application of functional failure concept in evaluation of risk-informed safety margins using as illustrative example the case of CANDU reactors shutdown systems effectiveness. A risk-informed formulation is first introduced for estimation of a reasonable limit for the functional failure probability using a Swiss cheese model. It is concluded that more research is needed in this area and a deterministic - probabilistic approach may be a reasonable intermediate step for evaluation of functional failure probability at the system level. The views expressed in this paper are those of the authors and do not necessarily reflect those of CNSC, or any part thereof. (authors)

Architecture for Integrated Medical Model Dynamic Probabilistic Risk Assessment

Science.gov (United States)

Jaworske, D. A.; Myers, J. G.; Goodenow, D.; Young, M.; Arellano, J. D.

2016-01-01

Probabilistic Risk Assessment (PRA) is a modeling tool used to predict potential outcomes of a complex system based on a statistical understanding of many initiating events. Utilizing a Monte Carlo method, thousands of instances of the model are considered and outcomes are collected. PRA is considered static, utilizing probabilities alone to calculate outcomes. Dynamic Probabilistic Risk Assessment (dPRA) is an advanced concept where modeling predicts the outcomes of a complex system based not only on the probabilities of many initiating events, but also on a progression of dependencies brought about by progressing down a time line. Events are placed in a single time line, adding each event to a queue, as managed by a planner. Progression down the time line is guided by rules, as managed by a scheduler. The recently developed Integrated Medical Model (IMM) summarizes astronaut health as governed by the probabilities of medical events and mitigation strategies. Managing the software architecture process provides a systematic means of creating, documenting, and communicating a software design early in the development process. The software architecture process begins with establishing requirements and the design is then derived from the requirements.

SHARP - a framework for incorporating human interactions into PRA studies

International Nuclear Information System (INIS)

Hannaman, G.W.; Joksimovich, V.; Spurgin, A.J.; Worledge, D.H.

1985-01-01

Recently, increased attention has been given to understanding the role of humans in the safe operation of nuclear power plants. By virtue of the ability to combine equipment reliability with human reliability probabilistic risk assessment (PRA) technology was deemed capable of providing significant insights about the contributions of human interations in accident scenarios. EPRI recognized the need to strengthen the methodology for incorporating human interactions into PRAs as one element of their broad research program to improve the credibility of PRAs. This research project lead to the development and detailed description of SHARP (Systematic Human Application Reliability Procedure) in EPRI NP-3583. The objective of this paper is to illustrate the SHARP framework. This should help PRA analysts state more clearly their assumptions and approach no matter which human reliability assessment technique is used. SHARP includes a structure of seven analysis steps which can be formally or informally performed during PRAs. The seven steps are termed definition, screening, breakdown, representation, impact assessment, quantification, and documentation

Practical Application of PRA as an Integrated Design Tool for Space Systems

Science.gov (United States)

Kalia, Prince; Shi, Ying; Pair, Robin; Quaney, Virginia; Uhlenbrock, John

2013-01-01

This paper presents the application of the first comprehensive Probabilistic Risk Assessment (PRA) during the design phase of a joint NASA/NOAA weather satellite program, Geostationary Operational Environmental Satellite Series R (GOES-R). GOES-R is the next generation weather satellite primarily to help understand the weather and help save human lives. PRA has been used at NASA for Human Space Flight for many years. PRA was initially adopted and implemented in the operational phase of manned space flight programs and more recently for the next generation human space systems. Since its first use at NASA, PRA has become recognized throughout the Agency as a method of assessing complex mission risks as part of an overall approach to assuring safety and mission success throughout project lifecycles. PRA is now included as a requirement during the design phase of both NASA next generation manned space vehicles as well as for high priority robotic missions. The influence of PRA on GOES-R design and operation concepts are discussed in detail. The GOES-R PRA is unique at NASA for its early implementation. It also represents a pioneering effort to integrate risks from both Spacecraft (SC) and Ground Segment (GS) to fully assess the probability of achieving mission objectives. PRA analysts were actively involved in system engineering and design engineering to ensure that a comprehensive set of technical risks were correctly identified and properly understood from a design and operations perspective. The analysis included an assessment of SC hardware and software, SC fault management system, GS hardware and software, common cause failures, human error, natural hazards, solar weather and infrastructure (such as network and telecommunications failures, fire). PRA findings directly resulted in design changes to reduce SC risk from micro-meteoroids. PRA results also led to design changes in several SC subsystems, e.g. propulsion, guidance, navigation and control (GNC

Uses of human reliability analysis probabilistic risk assessment results to resolve personnel performance issues that could affect safety

International Nuclear Information System (INIS)

O'Brien, J.N.; Spettell, C.M.

1985-10-01

This report is the first in a series which documents research aimed at improving the usefulness of Probabilistic Risk Assessment (PRA) results in addressing human risk issues. This first report describes the results of an assessment of how well currently available PRA data addresses human risk issues of current concern to NRC. Findings indicate that PRA data could be far more useful in addressing human risk issues with modification of the development process and documentation structure of PRAs. In addition, information from non-PRA sources could be integrated with PRA data to address many other issues. 12 tabs

MATILDA: A Military Laser Range Safety Tool Based on Probabilistic Risk Assessment (PRA) Techniques

Science.gov (United States)

2014-08-01

3 2.1 UK Need for a PRA-Based Approach ............................................................... 3 2.2 A Risk-Based Approach to...Figure 6: MATILDA Coordinate Transformations ....................................................... 22  Figure 7: Geocentric and MICS Coordinates...Star-Shaped Condition ................................................................................. 27  Figure 11: Points of Closest Approach

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

Review of cause-based decision tree approach for the development of domestic standard human reliability analysis procedure in low power/shutdown operation probabilistic safety assessment

International Nuclear Information System (INIS)

Kang, D. I.; Jung, W. D.

2003-01-01

We review the Cause-Based Decision Tree (CBDT) approach to decide whether we incorporate it or not for the development of domestic standard Human Reliability Analysis (HRA) procedure in low power/shutdown operation Probabilistic Safety Assessment (PSA). In this paper, we introduce the cause based decision tree approach, quantify human errors using it, and identify merits and demerits of it in comparision with previously used THERP. The review results show that it is difficult to incorporate the CBDT method for the development of domestic standard HRA procedure in low power/shutdown PSA because the CBDT method need for the subjective judgment of HRA analyst like as THERP. However, it is expected that the incorporation of the CBDT method into the development of domestic standard HRA procedure only for the comparision of quantitative HRA results will relieve the burden of development of detailed HRA procedure and will help maintain consistent quantitative HRA results

Spatial interactions database development for effective probabilistic risk assessment

International Nuclear Information System (INIS)

Liming, J. K.; Dunn, R. F.

2008-01-01

In preparation for a subsequent probabilistic risk assessment (PRA) fire risk analysis update, the STP Nuclear Operating Company (STPNOC) is updating its spatial interactions database (SID). This work is being performed to support updating the spatial interactions analysis (SIA) initially performed for the original South Texas Project Electric Generating Station (STPEGS) probabilistic safely assessment (PSA) and updated in the STPEGS Level 2 PSA and IPE Report. S/A is a large-scope screening analysis performed for nuclear power plant PRA that serves as a prerequisite basis for more detailed location-dependent, hazard-spec analyses in the PRA, such as fire risk analysis, flooding risk analysis, etc. SIA is required to support the 'completeness' argument for the PRA scope. The objectives of the current SID development effort are to update the spatial interactions analysis data, to the greatest degree practical, to be consistent with the following: the as-built plant as of December 31, 2007 the in-effect STPNOC STPEGS Units 1 and 2 PRA the current technology and intent of NUREG/CR-6850 guidance for lire risk analysis database support the requirements for PRA SIA, including fire and flooding risk analysis, established by NRC Regulatory Guide 1.200 and the ASME PRA Standard (ASME RA-S-2002 updated through ASME RA-Sc-2007,) This paper presents the approach and methodology for state-of-the-art SID development and applications, including an overview of the SIA process for nuclear power plant PRA. The paper shows how current relational database technology and existing, conventional station information sources can be employed to collect, process, and analyze spatial interactions data for the plant in an effective and efficient manner to meet the often challenging requirements of industry guidelines and standards such as NUREG/CR-6850, NRC Regulatory Guide 1.200, and ASME RA-S-2002 (updated through ASME RA-Sc 2007). This paper includes tables and figures illustrating how SIA

International status of application of probabilistic risk analysis

International Nuclear Information System (INIS)

Cullingford, M.C.

1984-01-01

Probabilistic Risk Assessment (PRA) having been practised for about ten years and with more than twenty studies completed has reached a level of maturity such that the insights and other products derived from specific studies may be assessed. The first full-scale PRA studies were designed to develop the methodology and assess the overall risk from nuclear power. At present PRA is performed mostly for individual plants to identify core damage accident sequences and significant contributors to such sequences. More than 25 countries are utilizing insights from PRA, some from full-scale PRA studies and other countries by performing reliability analyses on safety systems identified as important contributors to one or more core melt sequences. Many Member States of the IAEA fall into one of three groups: those having (a) a large, (b) a medium number of reactor-years of operating experience and (c) those countries in the planning or feasibility study stages of a nuclear power programme. Of the many potential uses of PRA the decision areas of safety improvement by backfitting, development of operating procedures and as the basis of standards are felt to be important by countries of all three groups. The use of PRA in showing compliance with safety goals and for plant availability studies is held to be important only by those countries which have operating experience. The evolution of the PRA methodology has led to increased attention to quantification of uncertainties both in the probabilities and consequences. Although many products from performing a PRA do not rely upon overall risk numbers, increasing emphasis is being placed on the interpretation of uncertainties in risk numbers for use in decisions. International co-operation through exchange of information regarding experience with PRA methodology and its application to nuclear safety decisions will greatly enhance the widespread use of PRA. (author)

Impact of shutdown risk on risk-based assessment of technical specifications

International Nuclear Information System (INIS)

Deriot, S.

1992-10-01

This paper describes the current work performed by the Research and Development Division of EDF concerning risk-based assessment of Operating Technical Specifications (OTS). The current risk-based assessment of OTS at EDF is presented. Then, the level 1 Probabilistic Safety Assessment of unit 3 of the Paluel nuclear power station (called PSA 1300) is described. It is fully computerized and takes into account the risk in shutdown states. A case study is presented. It shows that the fact of considering shutdown risk suggests that the current OTS should be modified

Probabilistic safety assessment of the Fugen NPS

International Nuclear Information System (INIS)

Sotsu, Masutake; Iguchi, Yukihiro; Mizuno, Kouichi; Sato, Shinichirou; Shimizu, Miwako

1999-01-01

We performed a probabilistic safety assessment (PSA) on the Fugen NPS. The main topic of assessment was internal factors. We assessment core damage frequency (level 1 PSA) and containment damage frequency (level 2 PSA) during rated operation, and core damage frequency during shutdown (PSA during shutdowns). Our assessment showed that the core damage frequency of Fugen is well below the IAEA criteria for existing plants, that the conditional containment damage during shutdown is almost the target value of 0.1, and that the core damage frequency during shutdown is almost the same as that assessed during operation. These results confirm that the Fugen plant maintains a sufficient safety margin during shutdowns for regular inspections and for refueling. We developed and verified the effectiveness of an accident management plan incorporating the results of the assessment. (author)

Probabilistic risk assessment and its role in plant modifications

International Nuclear Information System (INIS)

Diederich, A.R.; McElroy, W.F.

1986-01-01

Electric Utilities today have a tool available to improve management's ability to evaluate nuclear power plant modifications (MODS). Probabilistic Risk Assessment (PRA), is a tool of choice since it can be applied to a specific situation such as MOD request review, bringing the perspectives of reliability, financial risk and consequences to the public in addition to the more rigid requirements like those associated with Quality Assurance or licensing criteria. The techniques used in the PRA process revolve about the creation and manipulation of Fault Trees and Event Trees, which are used to quantify the event sequences and reliability of plant systems in a logical framework. It is through these methods that chains of sequences, or events, are understood. The degree to which plant systems are modelled in the PRA can vary depending on resources and purpose. Philadelphia Elecrtric Company's PRA modelled ten (10) major systems but this number may increase during the application and updating process

Probabilistic risk assessment in the CPI

International Nuclear Information System (INIS)

Guymer, P.; Kaiser, G.D.; Mc Kelvey, T.C.; Hannaman, G.W.

1987-01-01

Probabilistic Risk Assessment (PRA) is a method of quantifying the frequency of occurrence and magnitude of the consequences of accidents in systems that contain hazardous materials such as toxic, flammable or explosive chemicals. The frequency and magnitude of the consequences are the basic elements in the definition of risk, often simply expressed as the product of frequency and magnitude, summed over all accident sequences. PRA is a mature technique that has been used to estimate risk for a number of industrial facilities: for example, the Canvey Island Petrochemical complex; the Port of Rotterdam; the Reactor Safety Study, the first study to put the risks associated with nuclear power into perspective; and the transportation of chlorine. PRA has been developed to a greater level of sophistication in the nuclear industry than in the chemical industry. In the nuclear area, its usefulness has been demonstrated by increased plant safety, engineering insights, and cost-saving recommendations. Data and methods have been developed to increase the level of realism of the treatment of operator actions in PRA studies. It can be stated generally that the same methods can be applied with equal success in the chemical industry. However, there are pitfalls into which the unwary nuclear-oriented PRA analyst may stumble if he does not bear in mind that there are significant differences between nuclear plants and chemical plants

Developing Probabilistic Safety Performance Margins for Unknown and Underappreciated Risks

Science.gov (United States)

Benjamin, Allan; Dezfuli, Homayoon; Everett, Chris

2015-01-01

Probabilistic safety requirements currently formulated or proposed for space systems, nuclear reactor systems, nuclear weapon systems, and other types of systems that have a low-probability potential for high-consequence accidents depend on showing that the probability of such accidents is below a specified safety threshold or goal. Verification of compliance depends heavily upon synthetic modeling techniques such as PRA. To determine whether or not a system meets its probabilistic requirements, it is necessary to consider whether there are significant risks that are not fully considered in the PRA either because they are not known at the time or because their importance is not fully understood. The ultimate objective is to establish a reasonable margin to account for the difference between known risks and actual risks in attempting to validate compliance with a probabilistic safety threshold or goal. In this paper, we examine data accumulated over the past 60 years from the space program, from nuclear reactor experience, from aircraft systems, and from human reliability experience to formulate guidelines for estimating probabilistic margins to account for risks that are initially unknown or underappreciated. The formulation includes a review of the safety literature to identify the principal causes of such risks.

A probabilistic risk assessment of Oconee Unit 3. Executive highlights 60

International Nuclear Information System (INIS)

1984-04-01

In 1980 the Nuclear Safety Analysis Center and Duke Power Co. joined in a project to provide the utility industry with a practical, useful example of the application of probabilistic risk assessment (PRA) methods. PRA is a structured analysis technique that accounts for all the failure possibilities that might conceivably lead to core damage. The technique uses probabilities as discriminators to determine which are most significant. The following were project objectives: to provide the host utility with an analytic model of the plant that describes and estimates the likelihood of failure combinations that could lead to core melt; to evaluate the risks to the plant and to the public; to improve utility capabilities in PRA methods and applications

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

Comparison between Canadian probabilistic safety assessment methods formulated by Atomic Energy of Canada limited and probabilistic risk assessment methods

International Nuclear Information System (INIS)

Shapiro, H.S.; Smith, J.E.

1989-01-01

The procedures used by Atomic Energy of Canada Limited (AECL) to perform probabilistic safety assessments (PRAs) differ somewhat from conventionally accepted probabilistic risk assessment (PRA) procedures used elsewhere. In Canada, PSA is used by AECL as an audit tool for an evolving design. The purpose is to assess the safety of the plant in engineering terms. Thus, the PSA procedures are geared toward providing engineering feedback so that necessary changes can be made to the design at an early stage, input can be made to operating procedures, and test and maintenance programs can be optimized in terms of costs. Most PRAs, by contrast, are performed in plants that are already built. Their main purpose is to establish the core melt frequency and the risk to the public due to core melt. Also, any design modification is very expensive. The differences in purpose and timing between PSA and PRA have resulted in differences in methodology and scope. The PSA procedures are used on all plants being designed by AECL

Developing and evaluating distributions for probabilistic human exposure assessments

Energy Technology Data Exchange (ETDEWEB)

Maddalena, Randy L.; McKone, Thomas E.

2002-08-01

This report describes research carried out at the Lawrence Berkeley National Laboratory (LBNL) to assist the U. S. Environmental Protection Agency (EPA) in developing a consistent yet flexible approach for evaluating the inputs to probabilistic risk assessments. The U.S. EPA Office of Emergency and Remedial Response (OERR) recently released Volume 3 Part A of Risk Assessment Guidance for Superfund (RAGS), as an update to the existing two-volume set of RAGS. The update provides policy and technical guidance on performing probabilistic risk assessment (PRA). Consequently, EPA risk managers and decision-makers need to review and evaluate the adequacy of PRAs for supporting regulatory decisions. A critical part of evaluating a PRA is the problem of evaluating or judging the adequacy of input distributions PRA. Although the overarching theme of this report is the need to improve the ease and consistency of the regulatory review process, the specific objectives are presented in two parts. The objective of Part 1 is to develop a consistent yet flexible process for evaluating distributions in a PRA by identifying the critical attributes of an exposure factor distribution and discussing how these attributes relate to the task-specific adequacy of the input. This objective is carried out with emphasis on the perspective of a risk manager or decision-maker. The proposed evaluation procedure provides consistency to the review process without a loss of flexibility. As a result, the approach described in Part 1 provides an opportunity to apply a single review framework for all EPA regions and yet provide the regional risk manager with the flexibility to deal with site- and case-specific issues in the PRA process. However, as the number of inputs to a PRA increases, so does the complexity of the process for calculating, communicating and managing risk. As a result, there is increasing effort required of both the risk professionals performing the analysis and the risk manager

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

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

International Nuclear Information System (INIS)

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

1994-06-01

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

Use of PRA techniques to optimize the design of the IRIS nuclear power plant

International Nuclear Information System (INIS)

Muhlheim, M.D.; Cletcher, J.W. II

2003-01-01

True design optimization of a plants inherent safety and performance characteristics results when a probabilistic risk assessment (PRA) is integrated with the plant-level design process. This is the approach being used throughout the design of the International Reactor Innovative and Secure (IRIS) nuclear power plant to maximize safety. A risk-based design optimization tool employing a 'one-button' architecture is being developed by the Oak Ridge National Laboratory to evaluate design changes; new modeling approaches, methods, or theories modeling uncertainties and completeness; physical assumptions; and data changes on component, cabinet, train, and system bases. Unlike current PRAs, the one-button architecture allows components, modules, and data to be interchanged at will with the probabilistic effect immediately apparent. Because all of the current and previous design, and data sets are available via the one-button architecture, the safety ramifications of design options are evaluated, feedback on design alternatives is immediate, and true optimization and understanding can be achieved. Thus, for the first time, PRA analysts and designers can easily determine the probabilistic implications of different design configurations and operating conditions in various combinations for the entire range of initiating events. The power of the one-button architecture becomes evident by the number of design alternatives that can be evaluated C11 component choices yielded 160 design alternatives. Surprisingly, the lessons learned can be counter-intuitive and significant. For example, one of the alternative designs for IRIS evaluated via this architecture revealed that because of common-cause failure probabilities, using the most reliable components actually decreased systems' reliability. (author)

Survey of probabilistic methods in safety and risk assessment for nuclear power plant licensing

International Nuclear Information System (INIS)

1984-04-01

After an overview about the goals and general methods of probabilistic approaches in nuclear safety the main features of probabilistic safety or risk assessment (PRA) methods are discussed. Mostly in practical applications not a full-fledged PRA is applied but rather various levels of analysis leading from unavailability assessment of systems over the more complex analysis of the probable core damage stages up to the assessment of the overall health effects on the total population from a certain practice. The various types of application are discussed in relation to their limitation and benefits for different stages of design or operation of nuclear power plants. This gives guidance for licensing staff to judge the usefulness of the various methods for their licensing decisions. Examples of the application of probabilistic methods in several countries are given. Two appendices on reliability analysis and on containment and consequence analysis provide some more details on these subjects. (author)

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

International Nuclear Information System (INIS)

Choi, Sun Yeong; Yang, Joo Eon

2009-01-01

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: · Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA · Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding · Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs

Insights from Guideline for Performance of Internal Flooding Probabilistic Risk Assessment (IFPRA)

Energy Technology Data Exchange (ETDEWEB)

Choi, Sun Yeong; Yang, Joo Eon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-10-15

An internal flooding (IF) risk assessment refers to the quantitative probabilistic safety assessment (PSA) treatment of flooding as a result of pipe and tank breaks inside the plants, as well as from other recognized flood sources. The industry consensus standard for Internal Events Probabilistic Risk Assessment (ASME-RA-Sb-2005) includes high-level and supporting technical requirements for developing internal flooding probabilistic risk assessment (IFPRA). This industry standard is endorsed in Regulatory Guide 1.200, Revision 1 as an acceptable approach for addressing the risk contribution from IF events for risk informed applications that require U.S. Nuclear Regulatory commission (NRC) approval. In 2006, EPRI published a draft report for IFPRA that addresses the requirements of the ASME PRA consensus standard and have made efforts to refine and update the final EPRI IFPRA guideline. Westinghouse has performed an IFPRA analysis for several nuclear power plants (NPPs), such as Watts Bar and Fort Calhoun, using the draft EPRI guidelines for development of an IFPRA. Proprietary methodologies have been developed to apply the EPRI guidelines. The objectives of the draft report for IFPRA guideline are to: {center_dot} Provide guidance for PSA practitioners in the performance of the elements of a PRA associated with internal flooding events consistent with the current state of the art for internal flooding PRA {center_dot} Provide guidance regarding acceptable approaches that is sufficient to meeting the requirements of the ASME PRA Standard associated with internal flooding {center_dot} Incorporate lessons learned in the performance of internal flooding PRAs including those identified as pilot applications of earlier drafts of this procedures guide The purpose of this paper is to present a vision for domestic nuclear power plants' IFPRA by comparing the method of the draft EPRI guidelines with the existing IFPRA method for domestic NPPs.

Nuclear power plant personnel errors in decision-making as an object of probabilistic risk assessment

International Nuclear Information System (INIS)

Reer, B.

1993-09-01

The integration of human error - also called man-machine system analysis (MMSA) - is an essential part of probabilistic risk assessment (PRA). A new method is presented which allows for a systematic and comprehensive PRA inclusions of decision-based errors due to conflicts or similarities. For the error identification procedure, new question techniques are developed. These errors are shown to be identified by looking at retroactions caused by subordinate goals as components of the overall safety relevant goal. New quantification methods for estimating situation-specific probabilities are developed. The factors conflict and similarity are operationalized in a way that allows their quantification based on informations which are usually available in PRA. The quantification procedure uses extrapolations and interpolations based on a poor set of data related to decision-based errors. Moreover, for passive errors in decision-making a completely new approach is presented where errors are quantified via a delay initiating the required action rather than via error probabilities. The practicability of this dynamic approach is demonstrated by a probabilistic analysis of the actions required during the total loss of feedwater event at the Davis-Besse plant 1985. The extensions of the ''classical'' PRA method developed in this work are applied to a MMSA of the decay heat removal (DHR) of the ''HTR-500''. Errors in decision-making - as potential roots of extraneous acts - are taken into account in a comprehensive and systematic manner. Five additional errors are identified. However, the probabilistic quantification results a nonsignificant increase of the DHR failure probability. (orig.) [de

Observations and insights from low power and shutdown studies: Grand Gulf Nuclear Power Plant during POS 5 of a refueling outage

International Nuclear Information System (INIS)

Whitehead, D.W.; Brown, T.D.; Forester, J.A.

1995-04-01

With the recent completion of the documentation of the results from the Grand Gulf Nuclear Power Plant Low Power and Shutdown (LP and S) project funded by the US Nuclear Regulatory Commission (NRC), detailed probabilistic risk assessment (PRA) information from a boiling water reactor (BWR) for a specific time period in LP and S conditions became available for examination. This report contains observations and insights extracted from an examination of: (1) results in the LP and S documentation; (2) the specific models and assumptions used in the LP and S analyses; (3) selected results from the full-power analysis; (4) the experience of the analysts who performed the original LP and S study; and (5) results from sensitivity calculations performed as part of this project to help determine the impact that model assumptions and data values had on the results from the original LP and S analysis. Specifically, this study makes observations on and develops insights from the estimates of core damage frequency and aggregate risk (early fatalities and total latent cancer fatalities) associated with operations during plant operational state (POS) 5 (i.e., basically cold shutdown as defined by Technical Specifications) during a refueling outage for traditional internal events. A discussion of similarities and differences between full power accidents and accidents during LP and S conditions is provided. As part of this discussion, core damage frequency and risks results are presented on a per hour and per calendar year basis, allowing alternative perspectives on both the core damage frequency and risk associated with these two operational states

Implementation of condition-dependent probabilistic risk assessment using surveillance data on passive components

International Nuclear Information System (INIS)

Lewandowski, Radoslaw; Denning, Richard; Aldemir, Tunc; Zhang, Jinsuo

2016-01-01

Highlights: • Condition-dependent probabilistic risk assessment (PRA). • Time-dependent characterization of plant-specific risk. • Containment bypass involving in secondary system piping and SCC in SG tubes. - Abstract: A great deal of surveillance data are collected for a nuclear power plant that reflect the changing condition of the plant as it ages. Although surveillance data are used to determine failure probabilities of active components for the plant’s probabilistic risk assessment (PRA) and to indicate the need for maintenance activities, they are not used in a structured manner to characterize the evolving risk of the plant. The present study explores the feasibility of using a condition-dependent PRA framework that takes a first principles approach to modeling the progression of degradation mechanisms to characterize evolving risk, periodically adapting the model to account for surveillance results. A case study is described involving a potential containment bypass accident sequence due to the progression of flow-accelerated corrosion in secondary system piping and stress corrosion cracking of steam generator tubes. In this sequence, a steam line break accompanied by failure to close of a main steam isolation valve results in depressurization of the steam generator and induces the rupture of one or more faulted steam generator tubes. The case study indicates that a condition-dependent PRA framework might be capable of providing early identification of degradation mechanisms important to plant risk.

Development of extreme rainfall PRA methodology for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2016-01-01

The objective of this study is to develop a probabilistic risk assessment (PRA) methodology for extreme rainfall with focusing on decay heat removal system of a sodium-cooled fast reactor. For the extreme rainfall, annual excess probability depending on the hazard intensity was statistically estimated based on meteorological data. To identify core damage sequence, event trees were developed by assuming scenarios that structures, systems and components (SSCs) important to safety are flooded with rainwater coming into the buildings through gaps in the doors and the SSCs fail when the level of rainwater on the ground or on the roof of the building becomes higher than thresholds of doors on first floor or on the roof during the rainfall. To estimate the failure probability of the SSCs, the level of water rise was estimated by comparing the difference between precipitation and drainage capacity. By combining annual excess probability and the failure probability of SSCs, the event trees led to quantification of core damage frequency, and therefore the PRA methodology for rainfall was developed. (author)

Shutdown and low-power operation at commercial nuclear power plants in the United States

International Nuclear Information System (INIS)

1993-09-01

The report contains the results of the NRC Staff's evaluation of shutdown and low-power operations at US commercial nuclear power plants. The report describes studies conducted by the staff in the following areas: Operating experience related to shutdown and low-power operations, probabilistic risk assessment of shutdown and low-power conditions and utility programs for planning and conducting activities during periods the plant is shut down. The report also documents evaluations of a number of technical issues regarding shutdown and low-power operations performed by the staff, including the principal findings and conclusions. Potential new regulatory requirements are discussed, as well as potential changes in NRC programs. A draft report was issued for comment in February 1992. This report is the final version and includes the responses to the comments along with the staff regulatory analysis of potential new requirements

Probabilistic aspects of risk analyses for hazardous facilities

International Nuclear Information System (INIS)

Morici, A.; Valeri, A.; Zaffiro, C.

1989-01-01

The work described in the paper discusses the aspects of the risk analysis concerned with the use of the probabilistic methodology, in order to see how this approach may affect the risk management of industrial hazardous facilities. To this purpose reference is done to the Probabilistic Risk Assessment (PRA) of nuclear power plants. The paper points out that even though the public aversion towards nuclear risks is still far from being removed, the probabilistic approach may provide a sound support to the decision making and authorization process for any industrial activity implying risk for the environment and the public health. It is opinion of the authors that the probabilistic techniques have been developed to a great level of sophistication in the nuclear industry and provided much more experience in this field than in others. For some particular areas of the nuclear applications, such as the plant reliability and the plant response to the accidents, these techniques have reached a sufficient level of maturity and so some results have been usefully taken as a measure of the safety level of the plant itself. The use of some limited safety goals is regarded as a relevant item of the nuclear licensing process. The paper claims that it is time now that these methods would be applied with equal success to other hazardous facilities, and makes some comparative consideration on the differences of these plants with nuclear power plants in order to understand the effect of these differences on the PRA results and on the use one intends to make with them. (author)

Application of probabilistic risk assessment methodology to fusion

International Nuclear Information System (INIS)

Piet, S.J.

1985-07-01

Probabilistic Risk Assessment (PRA) tools are applied to general fusion issues in a systematic way, generally qualitatively. The potential value of PRA to general fusion safety and economic issues is discussed. Several important design insights result: possible fault interactions must be minimized (decouple fault conditions), inherently safe designs must include provision for passively handling loss of site power and loss of coolant conditions, the reliability of the vacuum boundary appears vital to maximizing facility availabilty and minimizing safety risk, and economic analyses appear to be incomplete without consideration of potential availability loss from forced outrages. A modification to PRA formalism is introduced, called the fault interaction matrix. The fault interaction matrix contains information concerning what initial fault condition could lead to other fault conditions and with what frequency. Thus, the fault interaction matrix represents a way to present and measure the degree to which a designer has decoupled possible fault conditions in his design

Advanced neutron source reactor probabilistic flow blockage assessment

International Nuclear Information System (INIS)

Ramsey, C.T.

1995-08-01

The Phase I Level I Probabilistic Risk Assessment (PRA) of the conceptual design of the Advanced Neutron Source (ANS) Reactor identified core flow blockage as the most likely internal event leading to fuel damage. The flow blockage event frequency used in the original ANS PRA was based primarily on the flow blockage work done for the High Flux Isotope Reactor (HFIR) PRA. This report examines potential flow blockage scenarios and calculates an estimate of the likelihood of debris-induced fuel damage. The bulk of the report is based specifically on the conceptual design of ANS with a 93%-enriched, two-element core; insights to the impact of the proposed three-element core are examined in Sect. 5. In addition to providing a probability (uncertainty) distribution for the likelihood of core flow blockage, this ongoing effort will serve to indicate potential areas of concern to be focused on in the preliminary design for elimination or mitigation. It will also serve as a loose-parts management tool

Probabilistic risk analysis for the NASA space shuttle: a brief history and current work

International Nuclear Information System (INIS)

Pate-Cornell, Elisabeth; Dillon, Robin

2001-01-01

While NASA managers have always relied on risk analysis tools for the development and maintenance of space projects, quantitative and especially probabilistic techniques have been gaining acceptance in recent years. In some cases, the studies have been required, for example, to launch the Galileo spacecraft with plutonium fuel, but these successful applications have helped to demonstrate the benefits of these tools. This paper reviews the history of probabilistic risk analysis (PRA) by NASA for the space shuttle program and discusses the status of the on-going development of the Quantitative Risk Assessment System (QRAS) software that performs PRA. The goal is to have within NASA a tool that can be used when needed to update previous risk estimates and to assess the benefits of possible upgrades to the system

Review of the Diablo Canyon probabilistic risk assessment

International Nuclear Information System (INIS)

Bozoki, G.E.; Fitzpatrick, R.G.; Bohn, M.P.; Sabek, M.G.; Ravindra, M.K.; Johnson, J.J.

1994-08-01

This report details the review of the Diablo Canyon Probabilistic Risk Assessment (DCPRA). The study was performed under contract from the Probabilistic Risk Analysis Branch, Office of Nuclear Reactor Research, USNRC by Brookhaven National Laboratory. The DCPRA is a full scope Level I effort and although the review touched on all aspects of the PRA, the internal events and seismic events received the vast majority of the review effort. The report includes a number of independent systems analyses sensitivity studies, importance analyses as well as conclusions on the adequacy of the DCPRA for use in the Diablo Canyon Long Term Seismic Program

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions.

Science.gov (United States)

Kaufman, Leyla V; Wright, Mark G

2017-07-07

The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA) procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

Risk contribution from low power and shutdown of a pressurized water reactor

International Nuclear Information System (INIS)

Chu, T.L.; Pratt, W.T.

1997-01-01

During 1989 the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (a pressurized water reactor) and Grand Gulf (a boiling water reactor), were selected for study by Brookhaven National Laboratory and Sandia National Laboratories, respectively. The program objectives included assessing the risks of severe accidents initiated during plant operational states other than full power operation and comparing estimated core damage frequencies, important accident sequences, and other qualitative and quantitative results with full power accidents as assessed in NUREG-1150. The scope included a Level 3 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. A phased approach was used in Level 1. In Phase 1 the concept of plant operational states (POSs) was developed to provide a better representation of the plant as it transitions from power to non power operation. This included a coarse screening analysis of all POSs to identify vulnerable plant configurations, to characterize (on a high, medium, or low basis) potential frequencies of core damage accidents, and to provide a foundation for a detailed Phase 2 analysis. In Phase 2, selected POSs from both Grand Gulf and Surry were chosen for detailed analysis. For Grand Gulf, POS 5 (approximately Cold Shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage was selected. For Surry, three POSs representing the time the plant spends in mid loop operation were chosen for analysis. Level 1 and Level 2/3 results from the Surry analyses are presented

Review insights on the probabilistic risk assessment for the Limerick Generating Station

International Nuclear Information System (INIS)

1984-08-01

In recognition of the high population density around the Limerick Generating Station site and the proposed power level, the Philadelphia Electric Company, in response to NRC staff requests, conducted and submitted between March 1981 and November 1983 a probabilistic risk assessment (PRA) on internal event contributors and a severe accident risk assessment on external event contributors to assess risks posed by operation of the plant. The applicant has developed perspectives using PRA models on the safety profile of the Limerick plant and has altered the plant design to reduce accident vulnerabilities identified in these PRAs. The staff's review of the Limerick PRA has particularly emphasized the dominant accident sequences and the resulting insights into demonstration of compliance with regulatory requirments, unique design features and major plant vulnerabilities to assess the need for any additional measures to further improve the safety of the LGS. The staff's review insights and PRA safety review conclusions are presented in this report

Overview of methods for uncertainty analysis and sensitivity analysis in probabilistic risk assessment

International Nuclear Information System (INIS)

Iman, R.L.; Helton, J.C.

1985-01-01

Probabilistic Risk Assessment (PRA) is playing an increasingly important role in the nuclear reactor regulatory process. The assessment of uncertainties associated with PRA results is widely recognized as an important part of the analysis process. One of the major criticisms of the Reactor Safety Study was that its representation of uncertainty was inadequate. The desire for the capability to treat uncertainties with the MELCOR risk code being developed at Sandia National Laboratories is indicative of the current interest in this topic. However, as yet, uncertainty analysis and sensitivity analysis in the context of PRA is a relatively immature field. In this paper, available methods for uncertainty analysis and sensitivity analysis in a PRA are reviewed. This review first treats methods for use with individual components of a PRA and then considers how these methods could be combined in the performance of a complete PRA. In the context of this paper, the goal of uncertainty analysis is to measure the imprecision in PRA outcomes of interest, and the goal of sensitivity analysis is to identify the major contributors to this imprecision. There are a number of areas that must be considered in uncertainty analysis and sensitivity analysis for a PRA: (1) information, (2) systems analysis, (3) thermal-hydraulic phenomena/fission product behavior, (4) health and economic consequences, and (5) display of results. Each of these areas and the synthesis of them into a complete PRA are discussed

Integration of human reliability analysis into the probabilistic risk assessment process: Phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1984-10-01

A research program was initiated to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three-phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

Energy Technology Data Exchange (ETDEWEB)

Cetiner, Mustafa Sacit; none,; Flanagan, George F. [ORNL; Poore III, Willis P. [ORNL; Muhlheim, Michael David [ORNL

2014-07-30

An integrated simulation platform that couples probabilistic analysis-based tools with model-based simulation tools can provide valuable insights for reactive and proactive responses to plant operating conditions. The objective of this work is to demonstrate the benefits of a partial implementation of the Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Framework Specification through the coupling of advanced PRA capabilities and accurate multi-physics plant models. Coupling a probabilistic model with a multi-physics model will aid in design, operations, and safety by providing a more accurate understanding of plant behavior. This represents the first attempt at actually integrating these two types of analyses for a control system used for operations, on a faster than real-time basis. This report documents the development of the basic communication capability to exchange data with the probabilistic model using Reliability Workbench (RWB) and the multi-physics model using Dymola. The communication pathways from injecting a fault (i.e., failing a component) to the probabilistic and multi-physics models were successfully completed. This first version was tested with prototypic models represented in both RWB and Modelica. First, a simple event tree/fault tree (ET/FT) model was created to develop the software code to implement the communication capabilities between the dynamic-link library (dll) and RWB. A program, written in C#, successfully communicates faults to the probabilistic model through the dll. A systems model of the Advanced Liquid-Metal Reactor–Power Reactor Inherently Safe Module (ALMR-PRISM) design developed under another DOE project was upgraded using Dymola to include proper interfaces to allow data exchange with the control application (ConApp). A program, written in C+, successfully communicates faults to the multi-physics model. The results of the example simulation were successfully plotted.

Application of database management software to probabilistic risk assessment calculations

International Nuclear Information System (INIS)

Wyss, G.D.

1993-01-01

Probabilistic risk assessment (PRA) calculations require the management and processing of large amounts of information. This data normally falls into two general categories. For example, a commercial nuclear power plant PRA study makes use of plant blueprints and system schematics, formal plant safety analysis reports, incident reports, letters, memos, handwritten notes from plant visits, and even the analyst's ''engineering judgment''. This information must be documented and cross-referenced in order to properly execute and substantiate the models used in a PRA study. The first category is composed of raw data that is accumulated from equipment testing and operational experiences. These data describe the equipment, its service or testing conditions, its failure mode, and its performance history. The second category is composed of statistical distributions. These distributions can represent probabilities, frequencies, or values of important parameters that are not time-related. Probability and frequency distributions are often obtained by fitting raw data to an appropriate statistical distribution. Database management software is used to store both types of data so that it can be readily queried, manipulated, and archived. This paper provides an overview of the information models used for storing PRA data and illustrates the implementation of these models using examples from current PRA software packages

Probabilistic risk assessment support of emergency preparedness at the Savannah River Site

International Nuclear Information System (INIS)

O'Kula, K.R.; Baker, W.H.; Simpkins, A.A.; Taylor, R.P.; Wagner, K.C.; Amos, C.N.

1992-01-01

Integration of the Probabilistic Risk Assessment (PRA) for K Reactor operation into related technical areas at the Savannah River Site (SRS) includes coordination with several onsite organizations responsible for maintaining and upgrading emergency preparedness capabilities. Major functional categories of the PRA application are scenario development and source term algorithm enhancement. Insights and technologies from the SRS PRA have facilitated development of: (1) credible timelines for scenarios; (2) algorithms tied to plant instrumentation to provide best-estimate source terms for dose projection; and (3) expert-system logic models to implement informed counter-measures to assure onsite and offsite safety following accidental releases. The latter methodology, in particular, is readily transferable to other reactor and non-reactor facilities at SRS and represents a distinct advance relative to emergency preparedness capabilities elsewhere in the DOE complex

On the use of data and judgment in probabilistic risk and safety analysis

International Nuclear Information System (INIS)

Kaplan, S.

1986-01-01

This paper reviews the line of thought of a nuclear plant probabilistic risk analysis (PRA) identifying the points where data and judgement enter. At the ''bottom'' of the process, data and judgment are combined, using one and two stage Bayesian methods, to express what is known about the element of variables. Higher in the process, we see the use of judgment in identifying scenarios and developing almost models and specifying initiating event categories. Finally, we discuss the judgments involved in deciding to do a PRA and in applying the results. (orig.)

Methodology and application of surrogate plant PRA analysis to the Rancho Seco Power Plant: Final report

International Nuclear Information System (INIS)

Gore, B.F.; Huenefeld, J.C.

1987-07-01

This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein

Preliminary ATWS analysis for the IRIS PRA

International Nuclear Information System (INIS)

Maddalena Barra; Marco S Ghisu; David J Finnicum; Luca Oriani

2005-01-01

to be a non dominant contribution to the plant CDF. During this second phase of the IRIS PRA development, the original model has been updated with the insertion of a study of ATWS, with the purpose of confirming the assumptions previously made. Within this framework, the most significant ATWS sequences have been identified and a first set of analyses has been run. The thermal-hydraulic computer code RELAP5 is used for the accident analyses, using a conservative evaluation model based on the Westinghouse ATWS evaluation model used during the analyses that supported the development of the ATWS rule (10CFR50.62). The key figure of merit used in these ATWS analyses is the pressure peak resulting from the imbalance between the rates of energy deposition into the core and removal from the reactor coolant that could cause damage to RCS components necessary for safe plant shutdown. Core damage is assumed if the peak pressure exceeds a limiting value. To support the evaluation of the ATWS contribution to the CDF, analyses with different values of critical parameters (mainly the moderator temperature coefficient and the number of operative safety valves in the pressurizer) were performed. The use of these analyses on the ATWS event tree development and on the calculation of the ATWS to the plant CDF is discussed in this paper. (authors)

Revision of the AESJ Standard for Seismic Probabilistic Risk Assessment (PRA). Updating requirements based on the lessons learned from the Fukushima Dai-ichi NPP Accidents (3). Fragility evaluation and outline of the updated points

International Nuclear Information System (INIS)

Yamaguchi, Akira; Nakamura, Susumu; Mihara, Yoshinori

2014-01-01

Lessons learned from Great East Japan earthquake and other new findings had been accumulated on the fragility evaluation of buildings and components. And also new analysis and evaluation method had been proposed with the advancement of recent analysis and evaluation technology. These were reflected in revision of the AESJ Standard for Seismic Probabilistic Risk Assessment (PRA). Scope of the fragility evaluation were extended to all equipment on the site, severe accident management equipment including portable equipment and earthquake concomitant incident (such as tsunami) countermeasure equipment. This article described outlines of updating points of the fragility evaluation of the AESJ Standard for Seismic PRA; (1) requirements for seismic induced other risk evaluations such as fire, inundation and tsunami, (2) simulation technology based on recent findings such as three dimensional responses of buildings / structures and its effect on equipment, (3) requirements of the fragility evaluation for various failure mode of several equipment such as severe accident management equipment, fine failure mode of buildings / structures, failures of equipment related with earthquake concomitant incidents (embankment and seawall) and spent fuel pool, and (4) requirements for the fragility evaluation of aftershocks and soil deformation due to fault displacement. (T. Tanaka)

Scram reliability under seismic conditions at the Experimental Breeder Reactor II

International Nuclear Information System (INIS)

Roglans, J.; Wang, C.Y.; Hill, D.J.

1993-01-01

A Probabilistic Risk Assessment of the Experimental Breeder Reactor II has recently been completed. Seismic events are among the external initiating events included in the assessment. As part of the seismic PRA a detailed study has been performed of the ability to shutdown the reactor under seismic conditions. A comprehensive finite element model of the EBR-II control rod drive system has been used to analyze the control rod system response when subjected to input seismic accelerators. The results indicate the control rod drive system has a high seismic capacity. The estimated seismic fragility for the overall reactor shutdown system is dominated by the primary tank failure

Critical safety function guidelines for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented

Probabilistic risk assessment: Number 219

International Nuclear Information System (INIS)

Bari, R.A.

1985-01-01

This report describes a methodology for analyzing the safety of nuclear power plants. A historical overview of plants in the US is provided, and past, present, and future nuclear safety and risk assessment are discussed. A primer on nuclear power plants is provided with a discussion of pressurized water reactors (PWR) and boiling water reactors (BWR) and their operation and containment. Probabilistic Risk Assessment (PRA), utilizing both event-tree and fault-tree analysis, is discussed as a tool in reactor safety, decision making, and communications. (FI)

Probabilistic risk criteria and their application to nuclear chemical plant design

International Nuclear Information System (INIS)

Arthur, T.; Barnes, D.S.; Brown, M.L.; Taig, A.R.; Johnston, B.D.; Hayns, M.

1989-01-01

A nuclear chemical plant safety strategy is presented. The use of risk criteria in design is demonstrated by reference to a particular area of the plant. This involves the application of Probabilistic Risk Assessment (PRA) techniques. Computer programs developed by the UK Atomic Energy Authority (UKAEA) at its Safety and Reliability Directorate (SRD) are used toe valuate and analyze the resultant fault trees. the magnitude of releases are estimated and individual and societal risks determined. The paper concludes that the application of PRA to a nuclear chemical plant can be structured in such a way as to allow a designer to work to quantitative risk targets

Engineering aspects of probabilistic risk assessment

International Nuclear Information System (INIS)

vonHerrmann, J.L.; Wood, P.J.

1984-01-01

Over the last decade, the use of probabilistic risk assessment (PRA) in the nuclear industry has expanded significantly. In these analyses the probabilities of experiencing certain undesired events (for example, a plant accident which results in damage to the nuclear fuel) are estimated and the consequences of these events are evaluated in terms of some common measure. These probabilities and consequences are then combined to form a representation of the risk associated with the plant studied. In the relatively short history of probabilistic risk assessment of nuclear power plants, the primary motivation for these studies has been the quantitative assessment of public risk associated with a single plant or group of plants. Accordingly, the primary product of most PRAs performed to date has been a 'risk curve' in which the probability (or expected frequency) of exceeding a certain consequence level is plotted against that consequence. The most common goal of these assessments has been to demonstrate the 'acceptability' of the calculated risk by comparison of the resultant risk curve to risk curves associated with other plants or with other societal risks. Presented here are brief descriptions of some alternate applications of PRAs, a discussion of how these other applications compare or contrast with the currently popular uses of PRA, and a discussion of the relative benefits of each

Assessing Probabilistic Risk Assessment Approaches for Insect Biological Control Introductions

Directory of Open Access Journals (Sweden)

Leyla V. Kaufman

2017-07-01

Full Text Available The introduction of biological control agents to new environments requires host specificity tests to estimate potential non-target impacts of a prospective agent. Currently, the approach is conservative, and is based on physiological host ranges determined under captive rearing conditions, without consideration for ecological factors that may influence realized host range. We use historical data and current field data from introduced parasitoids that attack an endemic Lepidoptera species in Hawaii to validate a probabilistic risk assessment (PRA procedure for non-target impacts. We use data on known host range and habitat use in the place of origin of the parasitoids to determine whether contemporary levels of non-target parasitism could have been predicted using PRA. Our results show that reasonable predictions of potential non-target impacts may be made if comprehensive data are available from places of origin of biological control agents, but scant data produce poor predictions. Using apparent mortality data rather than marginal attack rate estimates in PRA resulted in over-estimates of predicted non-target impact. Incorporating ecological data into PRA models improved the predictive power of the risk assessments.

Use of probabilistic risk assessment in maintenance activities at Palo Verde

International Nuclear Information System (INIS)

Lindquist, R.C.; Pobst, D.S.

1993-01-01

Probabilistic risk assessment (PRA) is an important tool in addressing various maintenance activities. At the Palo Verde nuclear generating station (PVNGS), the PRA has been used in a variety of ways to support a wide and diverse selection of maintenance-related activities. For on-line or at-power maintenance, the PRA was used to evaluate combinations of maintenance activities possible with the 12-week or floating maintenance schedule. The maintenance schedule was evaluated to identify any higher risk, undesirable combinations of equipment outages, such as the sole steam-driven auxiliary feedwater pump and the same train emergency diesel generator. Table I is a sampling of the results from the maintenance schedule evaluation in terms of increase in conditional core damage frequency (CDF) above the base- line value due to maintenance on some important key safety systems and combinations thereof. The baseline CDF is 7.4 x 10 -7 per 72 h

Spatially Informed Plant PRA Models for Security Assessment

International Nuclear Information System (INIS)

Wheeler, Timothy A.; Thomas, Willard; Thornsbury, Eric

2006-01-01

Traditional risk models can be adapted to evaluate plant response for situations where plant systems and structures are intentionally damaged, such as from sabotage or terrorism. This paper describes a process by which traditional risk models can be spatially informed to analyze the effects of compound and widespread harsh environments through the use of 'damage footprints'. A 'damage footprint' is a spatial map of regions of the plant (zones) where equipment could be physically destroyed or disabled as a direct consequence of an intentional act. The use of 'damage footprints' requires that the basic events from the traditional probabilistic risk assessment (PRA) be spatially transformed so that the failure of individual components can be linked to the destruction of or damage to specific spatial zones within the plant. Given the nature of intentional acts, extensive modifications must be made to the risk models to account for the special nature of the 'initiating events' associated with deliberate adversary actions. Intentional acts might produce harsh environments that in turn could subject components and structures to one or more insults, such as structural, fire, flood, and/or vibration and shock damage. Furthermore, the potential for widespread damage from some of these insults requires an approach that addresses the impacts of these potentially severe insults even when they occur in locations distant from the actual physical location of a component or structure modeled in the traditional PRA. (authors)

Medical Updates Number 5 to the International Space Station Probability Risk Assessment (PRA) Model Using the Integrated Medical Model

Science.gov (United States)

Butler, Doug; Bauman, David; Johnson-Throop, Kathy

2011-01-01

The Integrated Medical Model (IMM) Project has been developing a probabilistic risk assessment tool, the IMM, to help evaluate in-flight crew health needs and impacts to the mission due to medical events. This package is a follow-up to a data package provided in June 2009. The IMM currently represents 83 medical conditions and associated ISS resources required to mitigate medical events. IMM end state forecasts relevant to the ISS PRA model include evacuation (EVAC) and loss of crew life (LOCL). The current version of the IMM provides the basis for the operational version of IMM expected in the January 2011 timeframe. The objectives of this data package are: 1. To provide a preliminary understanding of medical risk data used to update the ISS PRA Model. The IMM has had limited validation and an initial characterization of maturity has been completed using NASA STD 7009 Standard for Models and Simulation. The IMM has been internally validated by IMM personnel but has not been validated by an independent body external to the IMM Project. 2. To support a continued dialogue between the ISS PRA and IMM teams. To ensure accurate data interpretation, and that IMM output format and content meets the needs of the ISS Risk Management Office and ISS PRA Model, periodic discussions are anticipated between the risk teams. 3. To help assess the differences between the current ISS PRA and IMM medical risk forecasts of EVAC and LOCL. Follow-on activities are anticipated based on the differences between the current ISS PRA medical risk data and the latest medical risk data produced by IMM.

Probabilistic Risk Assessment Procedures Guide for NASA Managers and Practitioners (Second Edition)

Science.gov (United States)

Stamatelatos,Michael; Dezfuli, Homayoon; Apostolakis, George; Everline, Chester; Guarro, Sergio; Mathias, Donovan; Mosleh, Ali; Paulos, Todd; Riha, David; Smith, Curtis;

PRA studies: results, insights and applications

International Nuclear Information System (INIS)

Levine, S.; Stetson, F.T.

1983-01-01

This paper deals with Probalistic Risk Assessment (PRA) studies and their results. The PRA is a combination of logic structures and analytical techniques that can be used to estimate the likelihood and consequences of events that have not been observed because of their low frequency occurrence. At first attitudes concerning PRA reports were controversial principally because of their new techniques and complex multidisciplinary nature. However these attitudes changed following the accident at Three Mile Island in 1979. Many people after this event came to appreciate the risks associated with the operation of nuclear power plants, and since the TMI accident there has been a rapid expansion, in the use of PRA in the US and other countries. (NEA) [fr

77 FR 61446 - Proposed Revision Probabilistic Risk Assessment and Severe Accident Evaluation for New Reactors

Science.gov (United States)

2012-10-09

... Severe Accident Evaluation for New Reactors AGENCY: Nuclear Regulatory Commission. ACTION: Standard... its Standard Review Plan (SRP), Section 19.0, ``Probabilistic Risk Assessment and Severe Accident... assessment (PRA) information and severe accident assessments for new reactors submitted to support design...

Integration of human reliability analysis into the probabilistic risk assessment process: phase 1

International Nuclear Information System (INIS)

Bell, B.J.; Vickroy, S.C.

1985-01-01

The US Nuclear Regulatory Commission and Pacific Northwest Laboratory initiated a research program in 1984 to develop a testable set of analytical procedures for integrating human reliability analysis (HRA) into the probabilistic risk assessment (PRA) process to more adequately assess the overall impact of human performance on risk. In this three phase program, stand-alone HRA/PRA analytic procedures will be developed and field evaluated to provide improved methods, techniques, and models for applying quantitative and qualitative human error data which systematically integrate HRA principles, techniques, and analyses throughout the entire PRA process. Phase 1 of the program involved analysis of state-of-the-art PRAs to define the structures and processes currently in use in the industry. Phase 2 research will involve developing a new or revised PRA methodology which will enable more efficient regulation of the industry using quantitative or qualitative results of the PRA. Finally, Phase 3 will be to field test those procedures to assure that the results generated by the new methodologies will be usable and acceptable to the NRC. This paper briefly describes the first phase of the program and outlines the second

OVERVIEW OF THE SAPHIRE PROBABILISTIC RISK ANALYSIS SOFTWARE

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis L.; Wood, Ted; Knudsen, James; Ma, Zhegang

2016-10-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer (PC) running the Microsoft Windows operating system. SAPHIRE Version 8 is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). INL's primary role in this project is that of software developer and tester. However, INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users, who constitute a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. In this paper, we provide an overview of the current technical capabilities found in SAPHIRE Version 8, including the user interface and enhanced solving algorithms.

Risk-informed design of IRIS using a level-1 probabilistic risk assessment from its conceptual design phase

International Nuclear Information System (INIS)

Mizuno, Yuko; Ninokata, Hisashi; Finnicum, David J.

2005-01-01

In this study, a probabilistic risk assessment (PRA) for the International Reactor Innovative and Secure (IRIS) has been generated to address two key areas as a part of the effort for the pre-application licensing of the IRIS design. First, the IRIS PRA is supporting the evaluation of IRIS design by providing design insights as well as a solid risk basis for the pre-licensing evaluation of the IRIS design. Second, the current PRA task is beginning the preparation of the more complete PRA analyses and documentation that will be required for Design Certification. The initial IRIS PRA is an at-power, Level-1 PRA for internal events that focuses on the evaluation of the IRIS design features to support the risk-informed design of IRIS by application of the PRA insights and the risk information to the design. To accomplish the evaluation, a reasonably complete Level-1 PRA model has been developed. The use of PRA in the early stages of the design has allowed a selection of design and performance features and an optimization of the design of several systems to reduce the potential for events that could lead to core damage via both enhanced prevention and mitigation of challenges. As a result, the total core damage frequency for internal events for the IRIS design has been calculated as 1.2x10 -8 per year

Applications of Living Fire PRA models to Fire Protection Significance Determination Process in Taiwan

International Nuclear Information System (INIS)

De-Cheng, Chen; Chung-Kung, Lo; Tsu-Jen, Lin; Ching-Hui, Wu; Lin, James C.

2004-01-01

The living fire probabilistic risk assessment (PRA) models for all three operating nuclear power plants (NPPs) in Taiwan had been established in December 2000. In that study, a scenario-based PRA approach was adopted to systematically evaluate the fire and smoke hazards and associated risks. Using these fire PRA models developed, a risk-informed application project had also been completed in December 2002 for the evaluation of cable-tray fire-barrier wrapping exemption. This paper presents a new application of the fire PRA models to fire protection issues using the fire protection significance determination process (FP SDP). The fire protection issues studied may involve the selection of appropriate compensatory measures during the period when an automatic fire detection or suppression system in a safety-related fire zone becomes inoperable. The compensatory measure can either be a 24-hour fire watch or an hourly fire patrol. The living fire PRA models were used to estimate the increase in risk associated with the fire protection issue in terms of changes in core damage frequency (CDF) and large early release frequency (LERF). In compliance with SDP at-power and the acceptance guidelines specified in RG 1.174, the fire protection issues in question can be grouped into four categories; red, yellow, white and green, in accordance with the guidelines developed for FD SDP. A 24-hour fire watch is suggested only required for the yellow condition, while an hourly fire patrol may be adopted for the white condition. More limiting requirement is suggested for the red condition, but no special consideration is needed for the green condition. For the calculation of risk measures, risk impacts from any additional fire scenarios that may have been introduced, as well as more severe initiating events and fire damages that may accompany the fire protection issue should be considered carefully. Examples are presented in this paper to illustrate the evaluation process. (authors)

How to eliminate non-damaging earthquakes from the results of a probabilistic seismic hazard analysis (PSHA)-A comprehensive procedure with site-specific application

International Nuclear Information System (INIS)

Kluegel, Jens-Uwe

2009-01-01

The results of probabilistic seismic hazard analyses are frequently presented in terms of uniform hazard spectra or hazard curves with spectral accelerations as the output parameter. The calculation process is based on the evaluation of the probability of exceedance of specified acceleration levels without consideration of the damaging effects of the causative earthquakes. The same applies to the empirical attenuation equations for spectral accelerations used in PSHA models. This makes interpreting and using the results in engineering or risk applications difficult. Uniform hazard spectra and the associated hazard curves may contain a significant amount of contributions of weak, low-energy earthquakes not able to damage the seismically designed structures of nuclear power plants. For the development of realistic engineering designs and for realistic seismic probabilistic risk assessments (seismic PRA) it is necessary to remove the contribution of non-damaging earthquakes from the results of a PSHA. A detailed procedure for the elimination of non-damaging earthquakes based on the CAV (Cumulative Absolute Velocity)-filtering approach was developed and applied to the results of the large-scale PEGASOS probabilistic seismic hazard study for the site of the Goesgen nuclear power plant. The procedure considers the full scope of epistemic uncertainty and aleatory variability present in the PEGASOS study. It involves the development of a set of empirical correlations for CAV and the subsequent development of a composite distribution for the probability of exceedance of the damaging threshold of 0.16 gs. Additionally, a method was developed to measure the difference in the damaging effects of earthquakes of different strengths by the ratio of a power function of ARIAS-intensity or, in the ideal case, by the ratio of the square roots of the associated strong motion durations. The procedure was applied for the update of the Goesgen seismic PRA and for the confirmation of a

Critical safety function guidelines for experimental fusion facilities

International Nuclear Information System (INIS)

Cadwallader, L.C.

1989-01-01

As fusion experiments proceed toward deuterium-tritium operation, more attention is being given to public safety. This paper presents the four classes of functions that fusion experiments must provide to assure safe, stable shutdown and retention of radionuclides. These functions are referred to as critical safety functions (CSFs). Selecting CSFs is an important step in probabilistic risk assessment (PRA). An example of CSF selection and usage for the Compact Ignition Tokamak (CIT) is also presented. 10 refs., 6 figs

Analytical solutions of linked fault tree probabilistic risk assessments using binary decision diagrams with emphasis on nuclear safety applications

International Nuclear Information System (INIS)

Nusbaumer, O. P. M.

2007-01-01

This study is concerned with the quantification of Probabilistic Risk Assessment (PRA) using linked Fault Tree (FT) models. Probabilistic Risk assessment (PRA) of Nuclear Power Plants (NPPs) complements traditional deterministic analysis; it is widely recognized as a comprehensive and structured approach to identify accident scenarios and to derive numerical estimates of the associated risk levels. PRA models as found in the nuclear industry have evolved rapidly. Increasingly, they have been broadly applied to support numerous applications on various operational and regulatory matters. Regulatory bodies in many countries require that a PRA be performed for licensing purposes. PRA has reached the point where it can considerably influence the design and operation of nuclear power plants. However, most of the tools available for quantifying large PRA models are unable to produce analytically correct results. The algorithms of such quantifiers are designed to neglect sequences when their likelihood decreases below a predefined cutoff limit. In addition, the rare event approximation (e.g. Moivre's equation) is typically implemented for the first order, ignoring the success paths and the possibility that two or more events can occur simultaneously. This is only justified in assessments where the probabilities of the basic events are low. When the events in question are failures, the first order rare event approximation is always conservative, resulting in wrong interpretation of risk importance measures. Advanced NPP PRA models typically include human errors, common cause failure groups, seismic and phenomenological basic events, where the failure probabilities may approach unity, leading to questionable results. It is accepted that current quantification tools have reached their limits, and that new quantification techniques should be investigated. A novel approach using the mathematical concept of Binary Decision Diagram (BDD) is proposed to overcome these deficiencies

Probabilistic risk assessment methodology

International Nuclear Information System (INIS)

Shinaishin, M.A.

1988-06-01

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

Probabilistic risk assessment methodology

Energy Technology Data Exchange (ETDEWEB)

Shinaishin, M A

1988-06-15

The objective of this work is to provide the tools necessary for clear identification of: the purpose of a Probabilistic Risk Study, the bounds and depth of the study, the proper modeling techniques to be used, the failure modes contributing to the analysis, the classical and baysian approaches for manipulating data necessary for quantification, ways for treating uncertainties, and available computer codes that may be used in performing such probabilistic analysis. In addition, it provides the means for measuring the importance of a safety feature to maintaining a level of risk at a Nuclear Power Plant and the worth of optimizing a safety system in risk reduction. In applying these techniques so that they accommodate our national resources and needs it was felt that emphasis should be put on the system reliability analysis level of PRA. Objectives of such studies could include: comparing systems' designs of the various vendors in the bedding stage, and performing grid reliability and human performance analysis using national specific data. (author)

An overview of insights gained and lessons learned from U.S. plant-specific PRA studies

International Nuclear Information System (INIS)

Joksimovich, V.

1985-01-01

Probabilistic Risk Assessment (PRA) has been under development for over twenty years, but it has reached the level of widespread use only in the aftermath of the TMI accident. Over thirty PRAs have now been completed in the U.S. PRAs have been in the mainstream of many licensing decisions because the NRC recognizes that they provide independent and comprehensive plant safety audit. Some difficulties have been experienced leading to interpretive and intercomparison studies. Numerous global and plant-specific insights have been derived. A new application termed risk management is clearly emerging. (orig./HP)

Review of KSNP LPSD PSA model based of ANS LPSD PRA standard, rev.0

International Nuclear Information System (INIS)

Jang, S. C.; Park, J. H.; Kim, T. W.; Lim, H. G.; Yang, J. E.; Ha, J. J.

2004-02-01

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). One of these efforts is the Risk Informed/Performance-based Operation (RIPBO). This approach uses the risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In RIPBO, PSA quality is one of the most important things. The nuclear industry and regulatory body of U.S.A have developed a measure to evaluate the quality of PSA. NEI (Nuclear Energy Institute) has developed a guidance called 'NEI PRA Peer Review Guidance,' and NRC (Nuclear Regulatory Committee) and ASME have developed the 'PRA Standard.' In Korea, several projects are on going now, such as the extension of AOT/STI of RPS/ESFAS, Risk-informed In-service Inspection (RI-ISI). However, in Korea, there have been no attempts to evaluate the quality of PSA model itself. Therefore, we cannot be sure about the quality of PSA whether or not the present PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of PSA model quality is the basis for the RIPBO. In this report, we have evaluated the quality of PSA model at Low power and Shutdown operation model for Yongkwang 5 and 6 units based on the ANS LPSD PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of LPSD PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea

Comparison of Qualitative and Quantitative Risk Results for Shutdown Operation

International Nuclear Information System (INIS)

Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro; Hong, Sung Yull

2006-01-01

The Defense-In-Depth philosophy is a fundamental concept of nuclear safety. The objective of Defense-In- Depth (DID) evaluation is to assess the level of Defense- In-Depth maintained during the various plant maintenance activities. Especially for shutdown and outage operations, the Defense-In-Depth might be challenged due to the reduction in redundancy and diversity resulting from the maintenance. The qualitative defense-in-depth evaluation using deterministic trees such as SFAT (Safety Function Assessment Tree), can provide 'Safety' related information on the levels of defense-in-depth according to the plant configuration including the levels of redundancy and diversity. For the more reasonable color decision of SFAT, it is necessary to identify the risk impact of degradation of redundancy and diversity of mitigation systems. The probabilistic safety analysis for the shutdown status can provide risk information related on the degradation of redundancy and diversity level for the safety functions during outage. Insights from the both methods for the plant status can be the same or different. The results of DID approach and PSA for the shutdown state are compared in this paper

TRIGA forced shutdowns analysis

International Nuclear Information System (INIS)

Negut, Gheorghe; Laslau, Florica

2008-01-01

The need for improving the operation leads us to use new methods and strategies. Probabilistic safety assessments and statistical analysis provide insights useful for our reactor operation. This paper is dedicated to analysis of the forced shutdowns during the first reactor operation period, between 1980 to 1989. A forced shutdown data base was designed using data on forced shutdowns collected from the reactor operation logbooks. In order to sort out the forced shutdowns the records have the following fields: - current number, date, equipment failed, failure type (M for mechanical, E for electrical, D for irradiation device, U for human factor failure; - scram mode, SE for external scram, failure of reactor cooling circuits and/or irradiation devices, SR for reactor scram, exceeding of reactor nuclear parameters, SB for reactor scram by control rod drop, SM for manual scram required by the abnormal reactor status; - scram cause, giving more information on the forced shutdown. This data base was processed using DBase III. The data processing techniques are presented. To sort out the data, one of the criteria was the number of scrams per year, failure type, scram mode, etc. There are presented yearly scrams, total operation time in hours, total unavailable time, median unavailable time period, reactor availability A. There are given the formulae used to calculate the reactor operational parameters. There are shown the scrams per year in the 1980 to 1989 period, the reactor operation time per year, the reactor shutdown time per year and the operating time versus down time per year. Total number of scrams in the covered period was 643 which caused a reactor down time of 4282.25 hours. In a table the scrams as sorted on the failure type is shown. Summarising, this study emphasized some problems and difficulties which occurred during the TRIGA reactor operation at Pitesti. One main difficulty in creating this data base was the unstandardized scram record mode. Some times

Probabilistic safety assessment support for the maintenance rule at Duke Power Company

International Nuclear Information System (INIS)

Brewer, H. Duncan; Canady, Ken S.

1999-01-01

The Nuclear Regulatory Commission (NRC) published the Maintenance Rule on July 10, 1991 with an implementation date of July 10, 1996 . Maintenance rule implementation at the Duke Power Company has used probabilistic safety assessment (PSA) insights to help focus the monitoring of structures, systems and components (SSC) performance and to ensure that maintenance is effectively performed. This paper describes how the probabilistic risk assessment (PRA) group at the Duke Power Company provides support for the maintenance rule by performing the following tasks: (1) providing a member of the expert panel; (2) determining the risk-significant SSCs; (3) establishing SSC performance criteria for availability and reliability; (4) evaluating past performance and its impact on core damage risk as part of the periodic assessment; (5) providing input to the PRA matrix; (6) providing risk analyses of combinations of SSCs out of service; (7) providing support for the SENTINEL program; and (8) providing support for PSA training. These tasks are not simply tied to the initial implementation of the rule. The maintenance rule must be kept consistent with the current design and operation of the plant. This will require that the PRA models and the many PSA calculations performed to support the maintenance rule are kept up-to-date. Therefore, support of the maintenance rule will be one of the primary roles of the PSA group for the remainder of the life of the plant

Advanced Test Reactor probabilistic risk assessment methodology and results summary

International Nuclear Information System (INIS)

Eide, S.A.; Atkinson, S.A.; Thatcher, T.A.

1992-01-01

The Advanced Test Reactor (ATR) probabilistic risk assessment (PRA) Level 1 report documents a comprehensive and state-of-the-art study to establish and reduce the risk associated with operation of the ATR, expressed as a mean frequency of fuel damage. The ATR Level 1 PRA effort is unique and outstanding because of its consistent and state-of-the-art treatment of all facets of the risk study, its comprehensive and cost-effective risk reduction effort while the risk baseline was being established, and its thorough and comprehensive documentation. The PRA includes many improvements to the state-of-the-art, including the following: establishment of a comprehensive generic data base for component failures, treatment of initiating event frequencies given significant plant improvements in recent years, performance of efficient identification and screening of fire and flood events using code-assisted vital area analysis, identification and treatment of significant seismic-fire-flood-wind interactions, and modeling of large loss-of-coolant accidents (LOCAs) and experiment loop ruptures leading to direct damage of the ATR core. 18 refs

A comparison of integrated safety analysis and probabilistic risk assessment

International Nuclear Information System (INIS)

Damon, Dennis R.; Mattern, Kevin S.

2013-01-01

The U.S. Nuclear Regulatory Commission conducted a comparison of two standard tools for risk informing the regulatory process, namely, the Probabilistic Risk Assessment (PRA) and the Integrated Safety Analysis (ISA). PRA is a calculation of risk metrics, such as Large Early Release Frequency (LERF), and has been used to assess the safety of all commercial power reactors. ISA is an analysis required for fuel cycle facilities (FCFs) licensed to possess potentially critical quantities of special nuclear material. A PRA is usually more detailed and uses more refined models and data than an ISA, in order to obtain reasonable quantitative estimates of risk. PRA is considered fully quantitative, while most ISAs are typically only partially quantitative. The extension of PRA methodology to augment or supplant ISAs in FCFs has long been considered. However, fuel cycle facilities have a wide variety of possible accident consequences, rather than a few surrogates like LERF or core damage as used for reactors. It has been noted that a fuel cycle PRA could be used to better focus attention on the most risk-significant structures, systems, components, and operator actions. ISA and PRA both identify accident sequences; however, their treatment is quite different. ISA's identify accidents that lead to high or intermediate consequences, as defined in 10 Code of Federal Regulations (CFR) 70, and develop a set of Items Relied on For Safety (IROFS) to assure adherence to performance criteria. PRAs identify potential accident scenarios and estimate their frequency and consequences to obtain risk metrics. It is acceptable for ISAs to provide bounding evaluations of accident consequences and likelihoods in order to establish acceptable safety; but PRA applications usually require a reasonable quantitative estimate, and often obtain metrics of uncertainty. This paper provides the background, features, and methodology associated with the PRA and ISA. The differences between the

Analytical solutions of linked fault tree probabilistic risk assessments using binary decision diagrams with emphasis on nuclear safety applications[Dissertation 17286

Energy Technology Data Exchange (ETDEWEB)

Nusbaumer, O. P. M

2007-07-01

This study is concerned with the quantification of Probabilistic Risk Assessment (PRA) using linked Fault Tree (FT) models. Probabilistic Risk assessment (PRA) of Nuclear Power Plants (NPPs) complements traditional deterministic analysis; it is widely recognized as a comprehensive and structured approach to identify accident scenarios and to derive numerical estimates of the associated risk levels. PRA models as found in the nuclear industry have evolved rapidly. Increasingly, they have been broadly applied to support numerous applications on various operational and regulatory matters. Regulatory bodies in many countries require that a PRA be performed for licensing purposes. PRA has reached the point where it can considerably influence the design and operation of nuclear power plants. However, most of the tools available for quantifying large PRA models are unable to produce analytically correct results. The algorithms of such quantifiers are designed to neglect sequences when their likelihood decreases below a predefined cutoff limit. In addition, the rare event approximation (e.g. Moivre's equation) is typically implemented for the first order, ignoring the success paths and the possibility that two or more events can occur simultaneously. This is only justified in assessments where the probabilities of the basic events are low. When the events in question are failures, the first order rare event approximation is always conservative, resulting in wrong interpretation of risk importance measures. Advanced NPP PRA models typically include human errors, common cause failure groups, seismic and phenomenological basic events, where the failure probabilities may approach unity, leading to questionable results. It is accepted that current quantification tools have reached their limits, and that new quantification techniques should be investigated. A novel approach using the mathematical concept of Binary Decision Diagram (BDD) is proposed to overcome these

Application of sensitivity analysis in nuclear power plant probabilistic risk assessment studies

International Nuclear Information System (INIS)

Hirschberg, S.; Knochenhauer, M.

1986-01-01

Nuclear power plant probabilistic risk assessment (PRA) studies utilise many models, simplifications and assumptions. Also subjective judgement is widely applied due to lack of actual data. This results in significant uncertainties. Three general types of uncertainties have been identified: (1) parameter uncertainties, (2) modelling uncertainties, and (3) completeness uncertainties. The significance of some of the modelling assumptions and simplifications cannot be investigated by assignment and propagation of parameter uncertainties. In such cases the impact of different options may (and should) be studied by performing sensitivity analyses, which concentrate on the most critical elements. This paper describes several items suitable for close examination by means of application of sensitivity analysis, when performing a level 1 PRA. Sensitivity analyses are performed with respect to: (1) boundary conditions (success criteria, credit for non-safety systems, degree of detail in modelling of support functions), (2) operator actions, (3) treatment of common cause failures (CCFs). The items of main interest are continuously identified in the course of performing a PRA study, as well as by scrutinising the final results. The practical aspects of sensitivity analysis are illustrated by several applications from a recent PRA study. The critical importance of modelling assumptions is also demonstrated by implementation of some modelling features from another level 1 PRA into the reference model. It is concluded that sensitivity analysis leads to insights important for analysts, reviewers and decision makers. (author)

Human reliability analysis in support of a level 1 PRA for Surry during midloop operations

International Nuclear Information System (INIS)

Lin, J.C.; Bley, D.C.; Chu, T.-L.

2004-01-01

The objectives of this Level 1 probabilistic risk assessment (PRA) are to evaluate the important accident sequences initiated during midloop operations and to compare the qualitative and quantitative results with those for accidents initiated during power operations. The primary types of human actions analyzed in this study involve the dynamic operator actions and recovery actions that take place during the accident sequence following an initiating event. Two parts of the human actions were analyzed: failure to diagnose and failure to perform the action. The scope of the Level 1 PRA for Surry during midloop operations includes internal, fire, and flood initiating events. The major categories of dynamic operator actions taken during the accident sequence following an initiating event are: providing makeup to the reactor coolant system (RCS), restoring residual heat removal (RHR) cooling, establishing steam generator reflux cooling, establishing primary feed and spill, establishing gravity feed from refueling water storage tank (RWST), establishing high pressure recirculation, establishing recirculation spray, and cross-connecting RWSTs. All categories are not applicable to all initiating events and all plant operating states (POS). (author)

Probabilistic risk analysis and fault trees: Initial discussion of application to identification of risk at a wellhead

Science.gov (United States)

Rodak, C.; Silliman, S.

2012-02-01

Wellhead protection is of critical importance for managing groundwater resources. While a number of previous authors have addressed questions related to uncertainties in advective capture zones, methods for addressing wellhead protection in the presence of uncertainty in the chemistry of groundwater contaminants, the relationship between land-use and contaminant sources, and the impact on health of the receiving population are limited. It is herein suggested that probabilistic risk analysis (PRA) combined with fault trees (FT) provides a structure whereby chemical transport can be combined with uncertainties in source, chemistry, and health impact to assess the probability of negative health outcomes in the population. As such, PRA-FT provides a new strategy for the identification of areas of probabilistically high human health risk. Application of this approach is demonstrated through a simplified case study involving flow to a well in an unconfined aquifer with heterogeneity in aquifer properties and contaminant sources.

Observations on PRA and its applications

International Nuclear Information System (INIS)

Yeh, Y.-C.; Shieh, S.-L.

2004-01-01

An overview on the experience of PRA and its prospective application in Taiwan's three nuclear power plants is presented. Through the PRA, plant design improvements are performed and several engineering findings are illuminated. The sensitivity study including the internal, seismic, and typhoon events are conducted to justify items that can significantly reduce core meltdown risk. Its resulted plant betterment plans are thus highlighted accordingly. For PRA application, a risk-based inspection program for allocating inspection human resources has been resulted following the importance ranking of each component. The developing risk-based regulation to rationalize technical specification and maintenance program will also be entailed. To enhance the accuracy of the PRA model and its reproducibility, several issues are considered to have high priority for improvement such as external event data and analyses, uncertainty, common mode failure, human reliability, and the relative component importance. Highlight of their significance along with some typical sensitivity analyses are discussed for further investigation. (author)

Human reliability assessment and probabilistic risk assessment

International Nuclear Information System (INIS)

Embrey, D.E.; Lucas, D.A.

1989-01-01

Human reliability assessment (HRA) is used within Probabilistic Risk Assessment (PRA) to identify the human errors (both omission and commission) which have a significant effect on the overall safety of the system and to quantify the probability of their occurrence. There exist a variey of HRA techniques and the selection of an appropriate one is often difficult. This paper reviews a number of available HRA techniques and discusses their strengths and weaknesses. The techniques reviewed include: decompositional methods, time-reliability curves and systematic expert judgement techniques. (orig.)

Insights into PRA methodologies

International Nuclear Information System (INIS)

Gallagher, D.; Lofgren, E.; Atefi, B.; Liner, R.; Blond, R.; Amico, P.

1984-08-01

Probabilistic Risk Assessments (PRAs) for six nuclear power plants were examined to gain insight into how the choice of analytical methods can affect the results of PRAs. The PRA sreflectope considered was limited to internally initiated accidents sequences through core melt. For twenty methodological topic areas, a baseline or minimal methodology was specified. The choice of methods for each topic in the six PRAs was characterized in terms of the incremental level of effort above the baseline. A higher level of effort generally reflects a higher level of detail or a higher degree of sophistication in the analytical approach to a particular topic area. The impact on results was measured in terms of how additional effort beyond the baseline level changed the relative importance and ordering of dominant accident sequences compared to what would have been observed had methods corresponding to the baseline level of effort been employed. This measure of impact is a more useful indicator of how methods affect perceptions of plant vulnerabilities than changes in core melt frequency would be. However, the change in core melt frequency was used as a secondary measure of impact for nine topics where availability of information permitted. Results are presented primarily in the form of effort-impact matrices for each of the twenty topic areas. A suggested effort-impact profile for future PRAs is presented

International Space Station End-of-Life Probabilistic Risk Assessment

Science.gov (United States)

Duncan, Gary W.

2014-01-01

The International Space Station (ISS) end-of-life (EOL) cycle is currently scheduled for 2020, although there are ongoing efforts to extend ISS life cycle through 2028. The EOL for the ISS will require deorbiting the ISS. This will be the largest manmade object ever to be de-orbited therefore safely deorbiting the station will be a very complex problem. This process is being planned by NASA and its international partners. Numerous factors will need to be considered to accomplish this such as target corridors, orbits, altitude, drag, maneuvering capabilities etc. The ISS EOL Probabilistic Risk Assessment (PRA) will play a part in this process by estimating the reliability of the hardware supplying the maneuvering capabilities. The PRA will model the probability of failure of the systems supplying and controlling the thrust needed to aid in the de-orbit maneuvering.

Uncertainty and sensitivity studies supporting the interpretation of the results of TVO I/II PRA

International Nuclear Information System (INIS)

Holmberg, J.

1992-01-01

A comprehensive Level 1 probabilistic risk assessment (PRA) has been performed for the TVO I/II nuclear power units. As a part of the PRA project, uncertainties of risk models and methods were systematically studied in order to describe them and to demonstrate their impact by way of results. The uncertainty study was divided into two phases: a qualitative and a quantitative study. The qualitative study contained identification of uncertainties and qualitative assessments of their importance. The PRA was introduced, and identified assumptions and uncertainties behind the models were documented. The most significant uncertainties were selected by importance measures or other judgements for further quantitative studies. The quantitative study included sensitivity studies and propagation of uncertainty ranges. In the sensitivity studies uncertain assumptions or parameters were varied in order to illustrate the sensitivity of the models. The propagation of the uncertainty ranges demonstrated the impact of the statistical uncertainties of the parameter values. The Monte Carlo method was used as a propagation method. The most significant uncertainties were those involved in modelling human interactions, dependences and common cause failures (CCFs), loss of coolant accident (LOCA) frequencies and pressure suppression. The qualitative mapping out of the uncertainty factors turned out to be useful in planning quantitative studies. It also served as internal review of the assumptions made in the PRA. The sensitivity studies were perhaps the most advantageous part of the quantitative study because they allowed individual analyses of the significance of uncertainty sources identified. The uncertainty study was found reasonable in systematically and critically assessing uncertainties in a risk analysis. The usefulness of this study depends on the decision maker (power company) since uncertainty studies are primarily carried out to support decision making when uncertainties are

Procedures for the elicitation of expert judgements in the probabilistic risk analysis of radioactive waste repositories: an overview

International Nuclear Information System (INIS)

Watson, S.R.

1992-01-01

In modelling the consequences of a radioactive waste repository using Probabilistic Risk Analysis, it is necessary to use the judgement of experts both in assessing probabilities subjectively, and in choosing suitable analytic frameworks. This report presents the literature on these topics, first discussing the meaning of probability in PRA, and then giving an extensive review of what is known about how to elicit probabilities from experts. The report then provides an overview of the less well developed field of how best to use expertise in the construction of models for PRA. (author)

Risk-Informed Safety Assurance and Probabilistic Assessment of Mission-Critical Software-Intensive Systems

Science.gov (United States)

Guarro, Sergio B.

2010-01-01

This report validates and documents the detailed features and practical application of the framework for software intensive digital systems risk assessment and risk-informed safety assurance presented in the NASA PRA Procedures Guide for Managers and Practitioner. This framework, called herein the "Context-based Software Risk Model" (CSRM), enables the assessment of the contribution of software and software-intensive digital systems to overall system risk, in a manner which is entirely compatible and integrated with the format of a "standard" Probabilistic Risk Assessment (PRA), as currently documented and applied for NASA missions and applications. The CSRM also provides a risk-informed path and criteria for conducting organized and systematic digital system and software testing so that, within this risk-informed paradigm, the achievement of a quantitatively defined level of safety and mission success assurance may be targeted and demonstrated. The framework is based on the concept of context-dependent software risk scenarios and on the modeling of such scenarios via the use of traditional PRA techniques - i.e., event trees and fault trees - in combination with more advanced modeling devices such as the Dynamic Flowgraph Methodology (DFM) or other dynamic logic-modeling representations. The scenarios can be synthesized and quantified in a conditional logic and probabilistic formulation. The application of the CSRM method documented in this report refers to the MiniAERCam system designed and developed by the NASA Johnson Space Center.

The role of PRA in the safety assessment of VVER Nuclear Power Plants in Ukraine

International Nuclear Information System (INIS)

Kot, C.

1999-01-01

Ukraine operates thirteen (13) Soviet-designed pressurized water reactors, VVERS. All Ukrainian plants are currently operating with annually renewable permits until they update their safety analysis reports (SARs), in accordance with new SAR content requirements issued in September 1995, by the Nuclear Regulatory Authority and the Government Nuclear Power Coordinating Committee of Ukraine. The requirements are in three major areas: design basis accident (DBA) analysis, probabilistic risk assessment (PRA), and beyond design-basis accident (BDBA) analysis. The last two requirements, on PRA and BDBA, are new, and the DBA requirements are an expanded version of the older SAR requirements. The US Department of Energy (USDOE), as part of its Soviet-Designed Reactor Safety activities, is providing assistance and technology transfer to Ukraine to support their nuclear power plants (NPPs) in developing a Western-type technical basis for the new SARs. USDOE sponsored In-Depth Safety Assessments (ISAs) are in progress at three pilot nuclear reactor units in Ukraine, South Ukraine Unit 1, Zaporizhzhya Unit 5, and Rivne Unit 1, and a follow-on study has been initiated at Khmenytskyy Unit 1. The ISA projects encompass most areas of plant safety evaluation, but the initial emphasis is on performing a detailed, plant-specific Level 1 Internal Events PRA. This allows the early definition of the plant risk profile, the identification of risk significant accident sequences and plant vulnerabilities and provides guidance for the remainder of the safety assessments

Validation of seismic probabilistic risk assessments of nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, B.

1994-01-01

A seismic probabilistic risk assessment (PRA) of a nuclear plant requires identification and information regarding the seismic hazard at the plant site, dominant accident sequences leading to core damage, and structure and equipment fragilities. Uncertainties are associated with each of these ingredients of a PRA. Sources of uncertainty due to seismic hazard and assumptions underlying the component fragility modeling may be significant contributors to uncertainty in estimates of core damage probability. Design and construction errors also may be important in some instances. When these uncertainties are propagated through the PRA, the frequency distribution of core damage probability may span three orders of magnitude or more. This large variability brings into question the credibility of PRA methods and the usefulness of insights to be gained from a PRA. The sensitivity of accident sequence probabilities and high-confidence, low probability of failure (HCLPF) plant fragilities to seismic hazard and fragility modeling assumptions was examined for three nuclear power plants. Mean accident sequence probabilities were found to be relatively insensitive (by a factor of two or less) to: uncertainty in the coefficient of variation (logarithmic standard deviation) describing inherent randomness in component fragility; truncation of lower tail of fragility; uncertainty in random (non-seismic) equipment failures (e.g., diesel generators); correlation between component capacities; and functional form of fragility family. On the other hand, the accident sequence probabilities, expressed in the form of a frequency distribution, are affected significantly by the seismic hazard modeling, including slopes of seismic hazard curves and likelihoods assigned to those curves

The development and application of an integrated radiological risk assessment procedure using time-dependent probabilistic risk analysis

International Nuclear Information System (INIS)

Laurens, J.M.; Thompson, B.G.J.; Sumerling, T.J.

1990-01-01

During the past decade, the UKDoE has funded the development of an integrated assessment procedure centred around probabilistic risk analysis (p.r.a.) using Monte Carlo simulation techniques to account for the effects of parameter value uncertainty, including those associated with temporal changes in the environment over a postclosure period of about one million years. The influence of these changes can now be incorporated explicitly into the p.r.a. simulator VANDAL (Variability ANalysis of Disposal ALternatives) briefly described here. Although a full statistically converged time-dependent p.r.a. will not be demonstrated until the current Dry Run 3 trial is complete, illustrative examples are given showing the ability of VANDAL to represent spatially complex groundwater and repository systems evolving under the influence of climatic change. 18 refs., 10 figs., 1 tab

Use of PRA in Shuttle Decision Making Process

Science.gov (United States)

Boyer, Roger L.; Hamlin, Teri L.

2010-01-01

How do you use PRA to support an operating program? This presentation will explore how the Shuttle Program Management has used the Shuttle PRA in its decision making process. It will reveal how the PRA has evolved from a tool used to evaluate Shuttle upgrades like Electric Auxiliary Power Unit (EAPU) to a tool that supports Flight Readiness Reviews (FRR) and real-time flight decisions. Specific examples of Shuttle Program decisions that have used the Shuttle PRA as input will be provided including how it was used in the Hubble Space Telescope (HST) manifest decision. It will discuss the importance of providing management with a clear presentation of the analysis, applicable assumptions and limitations, along with estimates of the uncertainty. This presentation will show how the use of PRA by the Shuttle Program has evolved overtime and how it has been used in the decision making process providing specific examples.

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1994-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Implications of an HRA framework for quantifying human acts of commission and dependency: Development of a methodology for conducting an integrated HRA/PRA

International Nuclear Information System (INIS)

Barriere, M.T.; Luckas, W.J.; Brown, W.S.; Cooper, S.E.; Wreathall, J.; Bley, D.C.

1993-01-01

To support the development of a refined human reliability analysis (HRA) framework, to address identified HRA user needs and improve HRA modeling, unique aspects of human performance have been identified from an analysis of actual plant-specific events. Through the use of the refined framework, relationships between the following HRA, human factors and probabilistic risk assessment (PRA) elements were described: the PRA model, plant states, plant conditions, PRA basic events, unsafe human actions, error mechanisms, and performance shaping factors (PSFs). The event analyses performed in the context of the refined HRA framework, identified the need for new HRA methods that are capable of: evaluating a range of different error mechanisms (e.g., slips as well as mistakes); addressing errors of commission (EOCs) and dependencies between human actions; and incorporating the influence of plant conditions and multiple PSFs on human actions. This report discusses the results of the assessment of user needs, the refinement of the existing HRA framework, as well as, the current status on EOCs, and human dependencies

Probabilistic analysis of fires in nuclear plants

International Nuclear Information System (INIS)

Unione, A.; Teichmann, T.

1985-01-01

The aim of this paper is to describe a multilevel (i.e., staged) probabilistic analysis of fire risks in nuclear plants (as part of a general PRA) which maximizes the benefits of the FRA (fire risk assessment) in a cost effective way. The approach uses several stages of screening, physical modeling of clearly dominant risk contributors, searches for direct (e.g., equipment dependences) and secondary (e.g., fire induced internal flooding) interactions, and relies on lessons learned and available data from and surrogate FRAs. The general methodology is outlined. 6 figs., 10 tabs

Probabilistic Design Analysis (PDA) Approach to Determine the Probability of Cross-System Failures for a Space Launch Vehicle

Science.gov (United States)

Shih, Ann T.; Lo, Yunnhon; Ward, Natalie C.

2010-01-01

Quantifying the probability of significant launch vehicle failure scenarios for a given design, while still in the design process, is critical to mission success and to the safety of the astronauts. Probabilistic risk assessment (PRA) is chosen from many system safety and reliability tools to verify the loss of mission (LOM) and loss of crew (LOC) requirements set by the NASA Program Office. To support the integrated vehicle PRA, probabilistic design analysis (PDA) models are developed by using vehicle design and operation data to better quantify failure probabilities and to better understand the characteristics of a failure and its outcome. This PDA approach uses a physics-based model to describe the system behavior and response for a given failure scenario. Each driving parameter in the model is treated as a random variable with a distribution function. Monte Carlo simulation is used to perform probabilistic calculations to statistically obtain the failure probability. Sensitivity analyses are performed to show how input parameters affect the predicted failure probability, providing insight for potential design improvements to mitigate the risk. The paper discusses the application of the PDA approach in determining the probability of failure for two scenarios from the NASA Ares I project

Space shuttle main propulsion pressurization system probabilistic risk assessment

International Nuclear Information System (INIS)

Plastiras, J.K.

1989-01-01

This paper reports that, in post-Challenger discussions with Congressional Committees and the National Research Council Risk Management Oversight Panel, criticism was levied against NASA because of the inability to prioritize the 1300+ single point failures. In the absence of a ranking it was difficult to determine where special effort was needed in failure evaluation, in design improvement, in management review of problems, and in flight readiness reviews. The belief was that the management system was overwhelmed by the quantity of critical hardware items that were on the Critical Items List (CIL) and that insufficient attention was paid to the items that required it. Congressional staff members from Congressman Markey's committee who have oversight responsibilities in the nuclear industry, and specifically over the nuclear power supplies for NASA's Galileo and Ulysses missions, felt very strongly that the addition of Probabilistic Risk Assessment (PRA) to the existing Failure Mode Effects Analysis/Hazard Analysis (FMEA/HA) methods was exceedingly important. Specifically, the Markey committee recognized that the inductive, qualitative component-oriented FMEA could be supplemented by the deductive, quantitative systems-oriented PRA. Furthermore, they felt that the PRA approach had matured to the extent that it could be used to assess risk, even with limited shuttle-specific failure data. NASA responded with arguments that the FMEA/HA had illuminated all significant failure modes satisfactorily and that no failure rate data base was available to support the PRA approach

Uncertainty estimation in nuclear power plant probabilistic safety assessment

International Nuclear Information System (INIS)

Guarro, S.B.; Cummings, G.E.

1989-01-01

Probabilistic Risk Assessment (PRA) was introduced in the nuclear industry and the nuclear regulatory process in 1975 with the publication of the Reactor Safety Study by the U.S. Nuclear Regulatory Commission. Almost fifteen years later, the state-of-the-art in this field has been expanded and sharpened in many areas, and about thirty-five plant-specific PRAs (Probabilistic Risk Assessments) have been performed by the nuclear utility companies or by the U.S. Nuclear Regulatory commission. Among the areas where the most evident progress has been made in PRA and PSA (Probabilistic Safety Assessment, as these studies are more commonly referred to in the international community outside the U.S.) is the development of a consistent framework for the identification of sources of uncertainty and the estimation of their magnitude as it impacts various risk measures. Techniques to propagate uncertainty in reliability data through the risk models and display its effect on the top level risk estimates were developed in the early PRAs. The Seismic Safety Margin Research Program (SSMRP) study was the first major risk study to develop an approach to deal explicitly with uncertainty in risk estimates introduced not only by uncertainty in component reliability data, but by the incomplete state of knowledge of the assessor(s) with regard to basic phenomena that may trigger and drive a severe accident. More recently NUREG-1150, another major study of reactor risk sponsored by the NRC, has expanded risk uncertainty estimation and analysis into the realm of model uncertainty related to the relatively poorly known post-core-melt phenomena which determine the behavior of the molten core and of the rector containment structures

Applications of probabilistic risk analysis in nuclear criticality safety design

International Nuclear Information System (INIS)

Chang, J.K.

1992-01-01

Many documents have been prepared that try to define the scope of the criticality analysis and that suggest adding probabilistic risk analysis (PRA) to the deterministic safety analysis. The report of the US Department of Energy (DOE) AL 5481.1B suggested that an accident is credible if the occurrence probability is >1 x 10 -6 /yr. The draft DOE 5480 safety analysis report suggested that safety analyses should include the application of methods such as deterministic safety analysis, risk assessment, reliability engineering, common-cause failure analysis, human reliability analysis, and human factor safety analysis techniques. The US Nuclear Regulatory Commission (NRC) report NRC SG830.110 suggested that major safety analysis methods should include but not be limited to risk assessment, reliability engineering, and human factor safety analysis. All of these suggestions have recommended including PRA in the traditional criticality analysis

MARATHON - a computer code for the probabilistic estimation of leak-before-break time in CANDU reactors

International Nuclear Information System (INIS)

Walker, J.R.

1990-02-01

The presence of high levels of moisture in the annulus gas system of a CANDU reactor indicates that a leaking crack may be present in a pressure tube. This will initiate the shutdown of the reactor to prevent the possibility of fuel channel damage. It is also desirable, however, to keep the reactor partially pressurized at hot shutdown for as long as it is necessary to unambiguously identify the leaking pressure tube. A premature full depressurization may cause an extended shutdown while the leaking tube is being located. However, fast fracture could occur during an excessively long hot shutdown period. A probabilistic methodology, together with an associated computer code (called MARATHON), has been developed to calculate the time from first leakage to unstable fracture in a probabilistic format. The methodology explicitly uses distributions of material properties and allows the risk associated with leak-before-break to be estimated. A model of the leak detection system is integrated into the methodology to calculate the time from leak detection to unstable fracture. The sensitivity of the risk to changing reactor conditions allows the optimization of reactor management after leak detection. In this report we describe the probabilistic model and give details of the quality assurance and verification of the MARATHON code. Examples of the use of MARATHON are given using preliminary material property distributions. These preliminary material property distributions indicate that the probability of unstable fracture is very low, and that ample time is available to locate the leaking tube

Validation needs of seismic probabilistic risk assessment (PRA) methods applied to nuclear power plants

International Nuclear Information System (INIS)

Kot, C.A.; Srinivasan, M.G.; Hsieh, B.J.

1985-01-01

An effort to validate seismic PRA methods is in progress. The work concentrates on the validation of plant response and fragility estimates through the use of test data and information from actual earthquake experience. Validation needs have been identified in the areas of soil-structure interaction, structural response and capacity, and equipment fragility. Of particular concern is the adequacy of linear methodology to predict nonlinear behavior. While many questions can be resolved through the judicious use of dynamic test data, other aspects can only be validated by means of input and response measurements during actual earthquakes. A number of past, ongoing, and planned testing programs which can provide useful validation data have been identified, and validation approaches for specific problems are being formulated

Probabilistic simulation applications to reliability assessments

International Nuclear Information System (INIS)

Miller, Ian; Nutt, Mark W.; Hill, Ralph S. III

2003-01-01

Probabilistic risk/reliability (PRA) analyses for engineered systems are conventionally based on fault-tree methods. These methods are mature and efficient, and are well suited to systems consisting of interacting components with known, low probabilities of failure. Even complex systems, such as nuclear power plants or aircraft, are modeled by the careful application of these approaches. However, for systems that may evolve in complex and nonlinear ways, and where the performance of components may be a sensitive function of the history of their working environments, fault-tree methods can be very demanding. This paper proposes an alternative method of evaluating such systems, based on probabilistic simulation using intelligent software objects to represent the components of such systems. Using a Monte Carlo approach, simulation models can be constructed from relatively simple interacting objects that capture the essential behavior of the components that they represent. Such models are capable of reflecting the complex behaviors of the systems that they represent in a natural and realistic way. (author)

Nuclear Regulatory Commission probabilistic risk assessment implementation program: A status report

International Nuclear Information System (INIS)

Rubin, M.P.; Caruso, M.A.

1996-01-01

The US Nuclear Regulatory Commission (NRC) is undertaking a number of activities intended to increase the consideration of risk significance in its decision processes and the effective use of risk-based technologies in its regulatory activities. Although the NRC is moving toward risk-informed regulation throughout its areas of responsibilities, this paper focuses primarily on those issues associated with reactor regulation. As the NRC completed significant milestones in its development of probabilistic risk assessment (PRA) methodology and gained considerable experience in the limited application of risk assessment to selected regulatory activities, it became evident that a much broader use of risk informed approaches offered advantages to both the NRC and the US commercial nuclear industry. This desire to enhance the use of risk assessment is driven by the clear belief that application of PRA methods will result in direct improvements in nuclear power plant operational safety from the perspective of both the regulator and the plant operator. The NRC believed that an overall policy on the use of PRA methods in nuclear regulatory activities should be established so that the many potential applications of PRA could be implemented in a consistent and predictable manner that would promote regulatory stability and efficiency. This paper describes the key activities that the NRC has undertaken to implement the initial stages of an integrated risk-informed regulatory framework

Issues related to structural aging in probabilistic risk assessment of nuclear power plants

International Nuclear Information System (INIS)

Ellingwood, Bruce R.

1998-01-01

Structural components and systems have an important safety function in nuclear power plants. Although they are essentially passive under normal operating conditions, they play a key role in mitigating the impact of extreme environmental events such as earthquakes, winds, fire and floods on plant safety. Moreover, the importance of structural components and systems in accident mitigation is amplified by common-cause effects. Reinforced concrete structural components and systems in NPPs are subject to a phenomenon known as aging, leading to time-dependent changes in strength and stiffness that may impact their ability to withstand various challenges during their service lives from operation, the environment and accidents. Time-dependent changes in structural properties as well as challenges to the system are random in nature. Accordingly, condition assessment of existing structures should be performed within a probabilistic framework. The mathematical formalism of a probabilistic risk assessment (PRA) provides a means for identifying aging structural components that may play a significant role in mitigating plant risk. Structural condition assessments supporting a decision regarding continued service can be rendered more efficient if guided by the logic of a PRA

A probabilistic risk assessment for field radiography based on expert judgment and opinion

International Nuclear Information System (INIS)

Jang, Han-Ki; Ryu, Hyung-Joon; Kim, Ji-Young; Lee, Jai-Ki; Cho, Kun-Woo

2011-01-01

A probabilistic approach was applied to assess radiation risk associated with the field radiography using gamma sources. The Delphi method based on the expert judgments and opinions was used in the process of characterization of parameters affecting risk, which are inevitably subject to large uncertainties. A mathematical approach applying the Bayesian inferences was employed for data processing to improve the Delphi results. This process consists of three phases: (1) setting prior distributions, (2) constructing the likelihood functions and (3) deriving the posterior distributions based on the likelihood functions. The approach for characterizing input parameters using the Bayesian inference is provided for improved risk estimates without intentional rejection of part of the data, which demonstrated utility of Bayesian updating of distributions of uncertain input parameters in PRA (Probabilistic Risk Assessment). The data analysis portion for PRA in field radiography is addressed for estimates of the parameters used to determine the frequencies and consequences of the various events modeled. In this study, radiological risks for the worker and the public member in the vicinity of the work place are estimated for field radiography system in Korea based on two-dimensional Monte Carlo Analysis (2D MCA). (author)

Human factors assessment in PRA using task analysis linked evaluation technique (TALENT)

International Nuclear Information System (INIS)

Wells, J.E.; Banks, W.W.

1990-01-01

Human error is a primary contributor to risk in complex high-reliability systems. A 1985 U.S. Nuclear Regulatory Commission (USNRC) study of licensee event reports (LERs) suggests that upwards of 65% of commercial nuclear system failures involve human error. Since then, the USNRC has initiated research to fully and properly integrate human errors into the probabilistic risk assessment (PRA) process. The resulting implementation procedure is known as the Task Analysis Linked Evaluation Technique (TALENT). As indicated, TALENT is a broad-based method for integrating human factors expertise into the PRA process. This process achieves results which: (1) provide more realistic estimates of the impact of human performance on nuclear power safety, (2) can be fully audited, (3) provide a firm technical base for equipment-centered and personnel-centered retrofit/redesign of plants enabling them to meet internally and externally imposed safety standards, and (4) yield human and hardware data capable of supporting inquiries into human performance issues that transcend the individual plant. The TALENT procedure is being field-tested to verify its effectiveness and utility. The objectives of the field-test are to examine (1) the operability of the process, (2) its acceptability to the users, and (3) its usefulness for achieving measurable improvements in the credibility of the analysis. The field-test will provide the information needed to enhance the TALENT process

Reliability and Probabilistic Risk Assessment - How They Play Together

Science.gov (United States)

Safie, Fayssal M.; Stutts, Richard G.; Zhaofeng, Huang

2015-01-01

PRA methodology is one of the probabilistic analysis methods that NASA brought from the nuclear industry to assess the risk of LOM, LOV and LOC for launch vehicles. PRA is a system scenario based risk assessment that uses a combination of fault trees, event trees, event sequence diagrams, and probability and statistical data to analyze the risk of a system, a process, or an activity. It is a process designed to answer three basic questions: What can go wrong? How likely is it? What is the severity of the degradation? Since 1986, NASA, along with industry partners, has conducted a number of PRA studies to predict the overall launch vehicles risks. Planning Research Corporation conducted the first of these studies in 1988. In 1995, Science Applications International Corporation (SAIC) conducted a comprehensive PRA study. In July 1996, NASA conducted a two-year study (October 1996 - September 1998) to develop a model that provided the overall Space Shuttle risk and estimates of risk changes due to proposed Space Shuttle upgrades. After the Columbia accident, NASA conducted a PRA on the Shuttle External Tank (ET) foam. This study was the most focused and extensive risk assessment that NASA has conducted in recent years. It used a dynamic, physics-based, integrated system analysis approach to understand the integrated system risk due to ET foam loss in flight. Most recently, a PRA for Ares I launch vehicle has been performed in support of the Constellation program. Reliability, on the other hand, addresses the loss of functions. In a broader sense, reliability engineering is a discipline that involves the application of engineering principles to the design and processing of products, both hardware and software, for meeting product reliability requirements or goals. It is a very broad design-support discipline. It has important interfaces with many other engineering disciplines. Reliability as a figure of merit (i.e. the metric) is the probability that an item will

Probabilistic risk assessment in the nuclear power industry

International Nuclear Information System (INIS)

Fullwood, R.R.; Hall, R.E.

1988-01-01

This book describes the more important improvements in risk assessment methodology developed over the last decade. The book covers the following areas - a general view of risk pertaining to nuclear power, mathematics necessary to understand the text, a concise overview of the light water reactors and their features for protecting the public, probabilities and consequences calculated to form risk assessment to the plant, and 34 applications of probabilistic risk assessment (PRA) in the power generation industry. There is a glossary of acronyms and unusual words and a list of references. (author)

Reliability modeling of Clinch River breeder reactor electrical shutdown systems

International Nuclear Information System (INIS)

Schatz, R.A.; Duetsch, K.L.

1974-01-01

The initial simulation of the probabilistic properties of the Clinch River Breeder Reactor Plant (CRBRP) electrical shutdown systems is described. A model of the reliability (and availability) of the systems is presented utilizing Success State and continuous-time, discrete state Markov modeling techniques as significant elements of an overall reliability assessment process capable of demonstrating the achievement of program goals. This model is examined for its sensitivity to safe/unsafe failure rates, sybsystem redundant configurations, test and repair intervals, monitoring by reactor operators; and the control exercised over system reliability by design modifications and the selection of system operating characteristics. (U.S.)

Management and Organization Influences in PRA

International Nuclear Information System (INIS)

Gertman, D.I.; Hallbert, B. P.; Blackman, H. S.

1998-01-01

The authors present a research program which aimed at increasing the quality of comprehensiveness of contemporary PRA (Probability Risk Assessment) by providing a tool that allows for incorporating M and O in PRA, at improving the quality of NRC assessments, at conducting research to support the risk informed regulation process, at identifying impact of management and organization, safety culture, workplace environment, down-sizing and deregulation on human performance and reliability

RAVEN: a GUI and an Artificial Intelligence Engine in a Dynamic PRA Framework

Energy Technology Data Exchange (ETDEWEB)

C. Rabiti; D. Mandelli; A. Alfonsi; J. Cogliati; R. Kinoshita; D. Gaston; R. Martineau; C. Curtis

2013-06-01

Increases in computational power and pressure for more accurate simulations and estimations of accident scenario consequences are driving the need for Dynamic Probabilistic Risk Assessment (PRA) [1] of very complex models. While more sophisticated algorithms and computational power address the back end of this challenge, the front end is still handled by engineers that need to extract meaningful information from the large amount of data and build these complex models. Compounding this problem is the difficulty in knowledge transfer and retention, and the increasing speed of software development. The above-described issues would have negatively impacted deployment of the new high fidelity plant simulator RELAP-7 (Reactor Excursion and Leak Analysis Program) at Idaho National Laboratory. Therefore, RAVEN that was initially focused to be the plant controller for RELAP-7 will help mitigate future RELAP-7 software engineering risks. In order to accomplish this task, Reactor Analysis and Virtual Control Environment (RAVEN) has been designed to provide an easy to use Graphical User Interface (GUI) for building plant models and to leverage artificial intelligence algorithms in order to reduce computational time, improve results, and help the user to identify the behavioral pattern of the Nuclear Power Plants (NPPs). In this paper we will present the GUI implementation and its current capability status. We will also introduce the support vector machine algorithms and show our evaluation of their potentiality in increasing the accuracy and reducing the computational costs of PRA analysis. In this evaluation we will refer to preliminary studies performed under the Risk Informed Safety Margins Characterization (RISMC) project of the Light Water Reactors Sustainability (LWRS) campaign [3]. RISMC simulation needs and algorithm testing are currently used as a guidance to prioritize RAVEN developments relevant to PRA.

Applications of PRA in nuclear criticality safety

International Nuclear Information System (INIS)

McLaughlin, T.P.

1992-01-01

Traditionally, criticality accident prevention at Los Alamos has been based on a thorough review and understanding of proposed operations of changes to operations, involving both process supervision and criticality safety staff. The outcome of this communication was usually an agreement, based on professional judgement, that certain accident sequences were credible and had to be reduced in likelihood either by administrative controls or by equipment design and others were not credible, and thus did not warrant expenditures to further reduce their likelihood. The extent of analysis and documentation was generally in proportion to the complexity of the operation but did not include quantified risk assessments. During the last three years nuclear criticality safety related Probabilistic Risk Assessments (PRAs) have been preformed on operations in two Los Alamos facilities. Both of these were conducted in order to better understand the cost/benefit aspects of PRA's as they apply to largely ''hands-on'' operations with fissile material for which human errors or equipment failures significant to criticality safety are both rare and unique. Based on these two applications and an appreciation of the historical criticality accident record (frequency and consequences) it is apparent that quantified risk assessments should be performed very selectively

Performing Probabilistic Risk Assessment Through RAVEN

Energy Technology Data Exchange (ETDEWEB)

A. Alfonsi; C. Rabiti; D. Mandelli; J. Cogliati; R. Kinoshita

2013-06-01

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data mining module

Two decades of PRA: What next?

International Nuclear Information System (INIS)

Rasmussen, N.C.

1992-01-01

Two decades ago, in the spring of 1972, the Reactor Safety Study was undertaken for the US Atomic Energy Commission (AEC). The goal of this study was to assess the risk to the public posed by the nuclear power plants operating in the US. Some three and one-half years later in October 1975, the study group issued its final report titled The Reactor Safety Study, also commonly known by its document number WASH 1400. Because it was issued at a time of heated public debate about nuclear safety, WASH 1400 received considerable critical review. By the late 1970s, as a result of the Lewis Report and the accident at Three Mile Island, the value of the WASH 1400 methodology was gradually recognized. A number of utilities undertook such studies of their own plants. The field of probabilistic risk assessment (PRA) developed from these efforts. Challenges remain. Among these are how to effectively communicate the results of the analysis. Just what does a probability of one in a million mean? Is there a de minimis probability - one so small that it can be ignored? How should society make decisions under substantial uncertainty? A number of these questions pose real challenges for the future

Development of margin assessment methodology of decay heat removal function against external hazards. (2) Tornado PRA methodology

International Nuclear Information System (INIS)

Nishino, Hiroyuki; Kurisaka, Kenichi; Yamano, Hidemasa

2014-01-01

Probabilistic Risk Assessment (PRA) for external events has been recognized as an important safety assessment method after the TEPCO's Fukushima Daiichi nuclear power station accident. The PRA should be performed not only for earthquake and tsunami which are especially key events in Japan, but also the PRA methodology should be developed for the other external hazards (e.g. tornado). In this study, the methodology was developed for Sodium-cooled Fast Reactors paying attention to that the ambient air is their final heat sink for removing decay heat under accident conditions. First, tornado hazard curve was estimated by using data recorded in Japan. Second, important structures and components for decay heat removal were identified and an event tree resulting in core damage was developed in terms of wind load and missiles (i.e. steel pipes, boards and cars) caused by a tornado. Main damage cause for important structures and components is the missiles and the tornado missiles that can reach those components and structures placed on high elevations were identified, and the failure probabilities of the components and structures against the tornado missiles were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or outtake in the decay heat removal system, and a probability of failure caused by the missile impacts. Finally, the event tree was quantified. As a result, the core damage frequency was enough lower than 10 -10 /ry. (author)

Assessment and presentation of uncertainties in probabilistic risk assessment: how should this be done

International Nuclear Information System (INIS)

Garlick, A.R.; Holloway, N.J.

1987-01-01

Despite continuing improvements in probabilistic risk assessment (PRA) techniques, PRA results, particularly those including degraded core analysis, will have maximum uncertainties of several orders of magnitude. This makes the expression of results, a matter no less important than their estimation. We put forward some ideas on the assessment and expression of highly uncertain quantities, such as probabilities of outcomes of a severe accident. These do not form a consistent set, but rather a number of alternative approaches aimed at stimulating discussion. These include non-probability expressions, such as fuzzy logic or Schafer's support and plausibility which abandon the purely probabilistic expression of risk for a more flexible type of expression, in which other types of measure are possible. The 'risk equivalent plant' concepts represent the opposite approach. Since uncertainty in a risk measure is in itself a form of risk, an attempt is made to define a 'risk equivalent' which is a risk with perfectly defined parameters, regarded (by means of suitable methods of judgement) as 'equally undesirable' with the actual plant. Some guidelines are given on the use of Bayesian methods in data-free or limited data situations. (author)

Individual plant examination and future PRA applications

International Nuclear Information System (INIS)

Monty, B.S.; Sursock, J.P.; Thierry, R.J.

1992-01-01

PRA is being used in many areas of plant operation as has been demonstrated in previous studies. With the U.S. NRC's emphasis on the use of risk to identify plant vulnerabilities and the development of plant specific PRA models for all plants, it is expected that the use of PRA will be expanded. Key areas where this is expected to occur include the development of risk-based Technical Specifications, risk management, and risk-centered maintenance programs. This paper focuses on the Individual Plant Examination requirement and the possible uses of risk-based methods in controlling plant operation to enhance plant safety and availability, and how the IPE requirement will potentially further this area of development. (orig./DG)

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

International Nuclear Information System (INIS)

Kmetyk, L.N.; Brown, T.D.

1995-03-01

To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP ampersand S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP ampersand S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP ampersand S configuration are given

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2

Energy Technology Data Exchange (ETDEWEB)

Kmetyk, L.N.; Brown, T.D. [Sandia National Labs., Albuquerque, NM (United States)

1995-03-01

To gain a better understanding of the risk significance of low power and shutdown modes of operation, the Office of Nuclear Regulatory Research at the NRC established programs to investigate the likelihood and severity of postulated accidents that could occur during low power and shutdown (LP&S) modes of operation at commercial nuclear power plants. To investigate the likelihood of severe core damage accidents during off power conditions, probabilistic risk assessments (PRAs) were performed for two nuclear plants: Unit 1 of the Grand Gulf Nuclear Station, which is a BWR-6 Mark III boiling water reactor (BWR), and Unit 1 of the Surry Power Station, which is a three-loop, subatmospheric, pressurized water reactor (PWR). The analysis of the BWR was conducted at Sandia National Laboratories while the analysis of the PWR was performed at Brookhaven National Laboratory. This multi-volume report presents and discusses the results of the BWR analysis. The subject of this part presents the deterministic code calculations, performed with the MELCOR code, that were used to support the development and quantification of the PRA models. The background for the work documented in this report is summarized, including how deterministic codes are used in PRAS, why the MELCOR code is used, what the capabilities and features of MELCOR are, and how the code has been used by others in the past. Brief descriptions of the Grand Gulf plant and its configuration during LP&S operation and of the MELCOR input model developed for the Grand Gulf plant in its LP&S configuration are given.

Shutdown problems in large tokamaks

International Nuclear Information System (INIS)

Weldon, D.M.

1978-01-01

Some of the problems connected with a normal shutdown at the end of the burn phase (soft shutdown) and with a shutdown caused by disruptive instability (hard shutdown) have been considered. For a soft shutdown a cursory literature search was undertaken and methods for controlling the thermal wall loading were listed. Because shutdown computer codes are not widespread, some of the differences between start-up codes and shutdown codes were discussed along with program changes needed to change a start-up code to a shutdown code. For a hard shutdown, the major problems are large induced voltages in the ohmic-heating and equilibrium-field coils and high first wall erosion. A literature search of plasma-wall interactions was carried out. Phenomena that occur at the plasma-wall interface can be quite complicated. For example, material evaporated from the wall can form a virtual limiter or shield protecting the wall from major damage. Thermal gradients that occur during the interaction can produce currents whose associated magnetic field also helps shield the wall

Seabrook Station Level 2 PRA Update to Include Accident Management

International Nuclear Information System (INIS)

Lutz, Robert; Lucci, Melissa; Kiper, Kenneth; Henry, Robert

2006-01-01

A ground-breaking study was recently completed as part of the Seabrook Level 2 PRA update. This study updates the post-core damage phenomena to be consistent with the most recent information and includes accident management activities that should be modeled in the Level 2 PRA. Overall, the result is a Level 2 PRA that fully meets the requirements of the ASME PRA Standard with respect to modeling accident management in the LERF assessment and NRC requirements in Regulatory Guide 1.174 for considering late containment failures. This technical paper deals only with the incorporation of operator actions into the Level 2 PRA based on a comprehensive study of the Seabrook Station accident response procedures and guidance. The paper describes the process used to identify the key operator actions that can influence the Level 2 PRA results and the development of success criteria for these key operator actions. This addresses a key requirement of the ASME PRA Standard for considering SAMG. An important benefit of this assessment was the identification of Seabrook specific accident management insights that can be fed back into the Seabrook Station accident management procedures and guidance or the training provided to plant personnel for these procedures and guidance. (authors)

Results of the AP600 advanced plant probabilistic risk assessment

International Nuclear Information System (INIS)

Bueter, T.; Sancaktar, S.; Freeland, J.

1997-01-01

The AP600 Probabilistic Risk Assessment (PRA) includes detailed models of the plant systems, including the containment and containment systems that would be used to mitigate the consequences of a severe accident. The AP600 PRA includes a level 1 analysis (core damage frequency), and a level 2 analysis (environmental consequences), an assessment of the plant vulnerability to accidents caused by fire or floods, and a seismic margins analysis. Numerous sensitivities are included in the AP600 PRA including one that assumes no credit for non-safety plant systems. The core damage frequency for the AP600 of 1.7E-07/year is small compared with other PRAs performed in the nuclear industry. The AP600 large release frequency of 1.8E-08/year is also small and shows the ability of the containment systems to prevent a large release should a severe accident occur. Analyses of potential consequences to the environment from a severe accident show that a release would be small, and that containment still provides significant protection 24 hours after an assumed accident. Sensitivity analyses show that plant risk (as measured by core damage frequency and large release frequency) is not sensitive to the reliability of operator actions. 6 refs., 1 fig., 1 tab

An integrated probabilistic risk analysis decision support methodology for systems with multiple state variables

International Nuclear Information System (INIS)

Sen, P.; Tan, John K.G.; Spencer, David

1999-01-01

Probabilistic risk analysis (PRA) methods have been proven to be valuable in risk and reliability analysis. However, a weak link seems to exist between methods for analysing risks and those for making rational decisions. The integrated decision support system (IDSS) methodology presented in this paper attempts to address this issue in a practical manner. In consists of three phases: a PRA phase, a risk sensitivity analysis (SA) phase and an optimisation phase, which are implemented through an integrated computer software system. In the risk analysis phase the problem is analysed by the Boolean representation method (BRM), a PRA method that can deal with systems with multiple state variables and feedback loops. In the second phase the results obtained from the BRM are utilised directly to perform importance and risk SA. In the third phase, the problem is formulated as a multiple objective decision making problem in the form of multiple objective reliability optimisation. An industrial example is included. The resultant solutions of a five objective reliability optimisation are presented, on the basis of which rational decision making can be explored

Nuclear reactor unit shutdown planning

International Nuclear Information System (INIS)

Gardais, J.P.

1994-01-01

In order to optimize the reactor maintenance shutdown efficiency and the reactor availability, an audit had been performed on the shutdown organization at EDF: management, skills, methods and experience feedback have been evaluated; several improvement paths have been identified: project management, introduction of shutdown management professionals, shutdown permanent industrialization, and experience feedback engineering

The tsunami probabilistic risk assessment of nuclear power plant (3). Outline of tsunami fragility analysis

International Nuclear Information System (INIS)

Mihara, Yoshinori

2012-01-01

Tsunami Probabilistic Risk Assessment (PRA) standard was issued in February 2012 by Standard Committee of Atomic Energy Society of Japan (AESJ). This article detailed tsunami fragility analysis, which calculated building and structure damage probability contributing core damage and consisted of five evaluation steps: (1) selection of evaluated element and damage mode, (2) selection of evaluation procedure, (3) evaluation of actual stiffness, (4) evaluation of actual response and (5) evaluation of fragility (damage probability and others). As an application example of the standard, calculation results of tsunami fragility analysis investigation by tsunami PRA subcommittee of AESJ were shown reflecting latest knowledge of damage state caused by wave force and others acted by tsunami from the 'off the Pacific Coast of Tohoku Earthquake'. (T. Tanaka)

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

International Nuclear Information System (INIS)

Coleman, Justin L.; Bolisetti, Chandu; Veeraraghavan, Swetha; Parisi, Carlo; Prescott, Steven R.; Gupta, Abhinav

2016-01-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

Multi-Hazard Advanced Seismic Probabilistic Risk Assessment Tools and Applications

Energy Technology Data Exchange (ETDEWEB)

Coleman, Justin L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bolisetti, Chandu [Idaho National Lab. (INL), Idaho Falls, ID (United States); Veeraraghavan, Swetha [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steven R. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gupta, Abhinav [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

Design of nuclear power plant (NPP) facilities to resist natural hazards has been a part of the regulatory process from the beginning of the NPP industry in the United States (US), but has evolved substantially over time. The original set of approaches and methods was entirely deterministic in nature and focused on a traditional engineering margins-based approach. However, over time probabilistic and risk-informed approaches were also developed and implemented in US Nuclear Regulatory Commission (NRC) guidance and regulation. A defense-in-depth framework has also been incorporated into US regulatory guidance over time. As a result, today, the US regulatory framework incorporates deterministic and probabilistic approaches for a range of different applications and for a range of natural hazard considerations. This framework will continue to evolve as a result of improved knowledge and newly identified regulatory needs and objectives, most notably in response to the NRC activities developed in response to the 2011 Fukushima accident in Japan. Although the US regulatory framework has continued to evolve over time, the tools, methods and data available to the US nuclear industry to meet the changing requirements have not kept pace. Notably, there is significant room for improvement in the tools and methods available for external event probabilistic risk assessment (PRA), which is the principal assessment approach used in risk-informed regulations and risk-informed decision-making applied to natural hazard assessment and design. This is particularly true if PRA is applied to natural hazards other than seismic loading. Development of a new set of tools and methods that incorporate current knowledge, modern best practice, and state-of-the-art computational resources would lead to more reliable assessment of facility risk and risk insights (e.g., the SSCs and accident sequences that are most risk-significant), with less uncertainty and reduced conservatisms.

Management of accidental scenarios involving the loss of RHRS under shutdown conditions

International Nuclear Information System (INIS)

Serradell, V.; Villanueva, J.F.; Martorell, S.; Carlos, S.; Pelayo, F.; Mendizabal, R.; Sol, I.

2009-01-01

Results from current Probabilistic Safety Assessment studies of Nuclear Power Plants show the importance of some risky scenarios with the plant at low power and shutdown conditions as compared to the accident scenarios with the plant operating at full power. Technical Specifications establish the Limiting Conditions for operation to assure the plant integrity in each Plant Operational State (POS). Moreover, the plant configuration may differ from the beginning to the end of a certain Plant Operational State, so the Limiting Conditions for Operation (LCO) established could be revised as, depending on the plant configuration, the transient evolution may be slightly different. For a PWR plant, one of the most risky accidental sequences in shutdown is the loss of the residual heat removal system, Using the information provided by the plant low power probabilistic safety analysis (LPSA), which should address the Limiting Conditions for Operation imposed by the current Technical Specification, two situations are distinguished: Main Reactor Cooling System (RCS) fully filled with water and RCS partially filled. In addition, while the primary system is partially filled in Cold Shutdown, two different plant configurations can be distinguished, which depend on the particular POS: RCS open and closed. For each case, the corresponding Technical Specification establishes the path to evacuate the residual heat generated. This paper explores the possibility of having alternative or complementary sources for heat removal others than the ones established in the Technical Specification. Especial attention is paid to the role of Steam Generators as an effective heat sink and the possibility of restart of the redundant RHR train. Such alternatives will influence LPSA implementation results. To perform this analysis the loss of the RHR system in a PWR plant has been simulated using RELAP-5 considering the plant in different plant operational states. One of the main results of this work

Application of FIVE methodology in probabilistic risk assessment (PRA) of fire events

International Nuclear Information System (INIS)

Lopez Garcia, F.J.; Suarez Alonso, J.; Fiolamengual, M.J.

1993-01-01

This paper reflects the experience acquired during the process of evaluation and updating of the fire analysis within the Cofrentes NPP PRA. It determines which points are the least precise, either because of their greater uncertainty or because of their excessive conservatism, as well as the subtasks which have involved a larger work load and could be simplified. These aspects are compared with the steps followed in methodology FIVE (Fire Vulnerability Evaluation Methodology) to assess whether application of this methodology would optimize the task, by making it more systematic and realistic and reducing uncertainties. On the one hand, the FIVE methodology does not have the scope sufficient to carry out a quantitative risk evaluation, but it can easily be complemented -without detriment to its systematic nature- by quantifying core damage in significant areas. On the other hand, certain issues such as definition of the fire growth software program which has to be used, are still not fully closed. Nevertheless, the conclusions derived from this assessment are satisfactory, since it is considered that this methodology would serve to unify the criteria and data of the analysis of fire-induced risks, providing a progressive screening method which would considerably simplify the task. (author)

Use of a probabilistic safety study in the design of the Italian reference PWR

International Nuclear Information System (INIS)

Richardson, D.C.; Russino, G.; Valentini, V.

1985-01-01

The intent of this paper is to provide a description of the experience gained in having performed a Probabilistic Safety Study (PSS) on the proposed Italian reference pressurized water reactor. The experience revealed that through careful application of probabilistic techniques, Probabilistic Risk Assessment (PRA) can be used as a tool to develop an optimum plant design in terms of safety and cost. Furthermore, the PSS can also be maintained as a living document and a tool to assess additional regulatory requirements that may be imposed during the construction and operational life of the plant. Through the use of flexible probabilistic techniques, the probabilistic safety model can provide a living safety assessment starting from the conceptual design and continuing through the construction, testing and operational phases. Moreover, the probabilistic safety model can be used during the operational phase of the plant as a method to evaluate the operational experience and identify potential problems before they occur. The experience, overall, provided additional insights into the various aspects of the plants design and operation that would not have been identified through the use of traditional safety evaluation techniques

Current and future applications of PRA in regulatory activities

Energy Technology Data Exchange (ETDEWEB)

Speis, T.P.; Murphy, J.A.; Cunningham, M.A. [Nuclear Regulatory Commission, Washington, DC (United States)] [and others

1995-04-01

Probabilistic Risk Assessments (PRAs) have proven valuable in providing the regulators, the nuclear plant operators, and the reactor designers insights into plant safety, reliability, design and operation. Both the NRC Commissioners and the staff have grown to appreciate the valuable contributions PRAs can have in the regulatory arena, though I will admit the existence of some tendencies for strict adherence to the deterministic approach within the agency and the public at large. Any call for change, particularly one involving a major adjustment in approach to the regulation of nuclear power, will meet with a certain degree of resistance and retrenchment. Change can appear threatening and can cause some to question whether the safety mission is being fulfilled. This skepticism is completely appropriate and is, in fact, essential to a proper transition towards risk and performance-based approaches. Our task in the Office of Nuclear Regulatory Research is to increase the PRA knowledge base within the agency and develop appropriate guidance and methods needed to support the transitioning process.

Simplified probabilistic risk assessment in fuel reprocessing

International Nuclear Information System (INIS)

Solbrig, C.W.

1993-01-01

An evaluation was made to determine if a backup mass tracking computer would significantly reduce the probability of criticality in the fuel reprocessing of the Integral Fast Reactor. Often tradeoff studies, such as this, must be made that would greatly benefit from a Probably Risk Assessment (PRA). The major benefits of a complete PRA can often be accrued with a Simplified Probabilistic Risk Assessment (SPRA). An SPRA was performed by selecting a representative fuel reprocessing operation (moving a piece of fuel) for analysis. It showed that the benefit of adding parallel computers was small compared to the benefit which could be obtained by adding parallelism to two computer input steps and two of the weighing operations. The probability of an incorrect material moves with the basic process is estimated to be 4 out of 100 moves. The actual values of the probability numbers are considered accurate to within an order of magnitude. The most useful result of developing the fault trees accrue from the ability to determine where significant improvements in the process can be made. By including the above mentioned parallelism, the error move rate can be reduced to 1 out of 1000

HTGR containment design options: an application of probabilistic risk assessment

International Nuclear Information System (INIS)

1977-08-01

Through the use of probabilistic risk assessment (PRA), it is possible to quantitatively evaluate the radiological risk associated with a given reactor design and to place such risk into perspective with alternative designs. The merits are discussed for several containment alternatives for the HTGR from the viewpoints of economics and licensability, as well as public risk. The quantification of cost savings and public risk indicates that presently acceptable public risk can be maintained and cost savings of $40 million can result from use of a vented confinement for the HTGR

CANDU passive shutdown systems

Energy Technology Data Exchange (ETDEWEB)

Hart, R S; Olmstead, R A [AECL CANDU, Sheridan Park Research Community, Mississauga, ON (Canada)

1996-12-01

CANDU incorporates two diverse, passive shutdown systems, independent of each other and from the reactor regulating system. Both shutdown systems function in the low pressure, low temperature, moderator which surrounds the fuel channels. The shutdown systems are functionally different, physically separate, and passive since the driving force for SDS1 is gravity and the driving force for SDS2 is stored energy. The physics of the reactor core itself ensures a degree of passive safety in that the relatively long prompt neutron generation time inherent in the design of CANDU reactors tend to retard power excursions and reduces the speed required for shutdown action, even for large postulated reactivity increases. All passive systems include a number of active components or initiators. Hence, an important aspect of passive systems is the inclusion of fail safe (activated by active component failure) operation. The mechanisms that achieve the fail safe action should be passive. Consequently the passive performance of the CANDU shutdown systems extends beyond their basic modes of operation to include fail safe operation based on natural phenomenon or stored energy. For example, loss of power to the SDS1 clutches results in the drop of the shutdown rods by gravity, loss of power or instrument air to the injection valves of SDS2 results in valve opening via spring action, and rigorous self checking of logic, data and timing by the shutdown systems computers assures a fail safe reactor trip through the collapse of a fluctuating magnetic field or the discharge of a capacitor. Event statistics from operating CANDU stations indicate a significant decrease in protection system faults that could lead to loss of production and elimination of protection system faults that could lead to loss of protection. This paper provides a comprehensive description of the passive shutdown systems employed by CANDU. (author). 4 figs, 3 tabs.

EPRI/NRC-RES fire PRA guide for nuclear power facilities. Volume 1, summary and overview

International Nuclear Information System (INIS)

2004-01-01

This report documents state-of-the-art methods, tools, and data for the conduct of a fire Probabilistic Risk Assessment (PRA) for a commercial nuclear power plant (NPP) application. The methods have been developed under the Fire Risk Re-quantification Study. This study was conducted as a joint activity between EPRI and the U. S. NRC Office of Nuclear Regulatory Research (RES) under the terms of an EPRI/RES Memorandum of Understanding (RS.1) and an accompanying Fire Research Addendum (RS.2). Industry participants supported demonstration analyses and provided peer review of this methodology. The documented methods are intended to support future applications of Fire PRA, including risk-informed regulatory applications. The documented method reflects state-of-the-art fire risk analysis approaches. The primary objective of the Fire Risk Study was to consolidate recent research and development activities into a single state-of-the-art fire PRA analysis methodology. Methodological issues raised in past fire risk analyses, including the Individual Plant Examination of External Events (IPEEE) fire analyses, have been addressed to the extent allowed by the current state-of-the-art and the overall project scope. Methodological debates were resolved through a consensus process between experts representing both EPRI and RES. The consensus process included a provision whereby each major party (EPRI and RES) could maintain differing technical positions if consensus could not be reached. No cases were encountered where this provision was invoked. While the primary objective of the project was to consolidate existing state-of-the-art methods, in many areas, the newly documented methods represent a significant advancement over previously documented methods. In several areas, this project has, in fact, developed new methods and approaches. Such advances typically relate to areas of past methodological debate.

The radioprotective effect of a new aminothiol (20-PRA)

International Nuclear Information System (INIS)

Dolabela, M.F.; Lopes, M.T.P.; Pereira, M.T.; Steffani, G.M.; Pilo-Veloso, D.; Salas, C.E.; Nelson, D.L.

1998-01-01

We examined the radioprotective effect of aminothiol 2-N-propylamine-cyclohexane thiol (20-PRA) on a human leukemic cell line (K562) following various radiation doses (5,7.5 and 20 Gy) using a source of 60 Co γ-rays. At 5 Gy and 1nM 20-PRA, a substantial protective effect (58%) was seen 24 h after irradiation, followed by a decrease at 48 h (11%). At the high radiation dose (20 Gy) a low protective effect was also seen (35%). In addition, the anti tumorigenic potential of 10 nM 20-PRA was shown by the inhibition of crown gall formation induced by Agrobacterium tumefaciens. The radioprotective potency of 20-PRA is 10 5- 10 6 times higher than that of the aminothiol WR-1065 (N(2-mercaptoethyl)-1,3-diamino propane) whose protective effect is in the 0.1 to 1.0 nM range. (author)

The radioprotective effect of a new aminothiol (20-PRA

Directory of Open Access Journals (Sweden)

M.F. Dolabela

1998-08-01

Full Text Available We examined the radioprotective effect of aminothiol 2-N-propylamine-cyclo-hexanethiol (20-PRA on a human leukemic cell line (K562 following various radiation doses (5, 7.5 and 20 Gy using a source of 60Co g-rays. At 5 Gy and 1 nM 20-PRA, a substantial protective effect (58% was seen 24 h after irradiation, followed by a decrease at 48 h (11%. At the high radiation dose (20 Gy a low protective effect was also seen (35%. In addition, the antitumorigenic potential of 10 nM 20-PRA was shown by the inhibition of crown gall formation induced by Agrobacterium tumefaciens. The radioprotective potency of 20-PRA is 105-106 times higher than that of the aminothiol WR-1065 (N-(2-mercaptoethyl-1,3-diaminopropane whose protective effect is in the 0.1 to 1.0 mM range.

PSA-operations synergism for the advanced test reactor shutdown operations PSA

International Nuclear Information System (INIS)

Atkinson, S.A.

1996-01-01

The Advanced Test Reactor (ATR) Probabilistic Safety Assessment (PSA) for shutdown operations, cask handling, and canal draining is a successful example of the importance of good PSA-operations synergism for achieving a realistic and accepted assessment of the risks and for achieving desired risk reduction and safety improvement in a best and cost-effective manner. The implementation of the agreed-upon upgrades and improvements resulted in the reductions of the estimated mean frequency for core or canal irradiated fuel uncovery events, a total reduction in risk by a factor of nearly 1000 to a very low and acceptable risk level for potentially severe events

Dynamical systems probabilistic risk assessment

Energy Technology Data Exchange (ETDEWEB)

Denman, Matthew R. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Ames, Arlo Leroy [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2014-03-01

Probabilistic Risk Assessment (PRA) is the primary tool used to risk-inform nuclear power regulatory and licensing activities. Risk-informed regulations are intended to reduce inherent conservatism in regulatory metrics (e.g., allowable operating conditions and technical specifications) which are built into the regulatory framework by quantifying both the total risk profile as well as the change in the risk profile caused by an event or action (e.g., in-service inspection procedures or power uprates). Dynamical Systems (DS) analysis has been used to understand unintended time-dependent feedbacks in both industrial and organizational settings. In dynamical systems analysis, feedback loops can be characterized and studied as a function of time to describe the changes to the reliability of plant Structures, Systems and Components (SSCs). While DS has been used in many subject areas, some even within the PRA community, it has not been applied toward creating long-time horizon, dynamic PRAs (with time scales ranging between days and decades depending upon the analysis). Understanding slowly developing dynamic effects, such as wear-out, on SSC reliabilities may be instrumental in ensuring a safely and reliably operating nuclear fleet. Improving the estimation of a plant's continuously changing risk profile will allow for more meaningful risk insights, greater stakeholder confidence in risk insights, and increased operational flexibility.

Clinical significance of determination of SAC/PRA value in patients with primary aldosteronism

International Nuclear Information System (INIS)

Li Liren; Dai Yaozong; Liu Jiumin

2003-01-01

Objective: To investigate the diagnostic significance of determining SAC/PRA valve in hyperaldosteronism. Methods: Plasma renin activity (PRA) and angiotensin (AT-II) as well as serum aldosterone contents were measured with RIA in 48 patients with primary aldosteronism and 30 controls. The SAC/PRA value was calculated. Results: Contents of PRA, AT-II and Aldo in blood of patients with primary aldosteronism were very significantly different from those in controls (p < 0.001) (PRA 0.14 ± 0.08 ng/ml/h vs 0.57 ± 0.08 ng/ml/h; AT-II 21.21 ± 7.55 ng/L vs 36.03 ± 6.11 ng/L; Aldo 1.07 ± 0.34 nmol/L vs 0.33 ± 0.04 nmol/L). Calculated SAC/PRA value was 913 ± 409 (normal upper limit 400). Conclusion: SAC/PRA value is an useful accessory diagnostic criterion for primary aldosteronism

Probabilistic Modeling of Wind Turbine Drivetrain Components

DEFF Research Database (Denmark)

Rafsanjani, Hesam Mirzaei

Wind energy is one of several energy sources in the world and a rapidly growing industry in the energy sector. When placed in offshore or onshore locations, wind turbines are exposed to wave excitations, highly dynamic wind loads and/or the wakes from other wind turbines. Therefore, most components...... in a wind turbine experience highly dynamic and time-varying loads. These components may fail due to wear or fatigue, and this can lead to unplanned shutdown repairs that are very costly. The design by deterministic methods using safety factors is generally unable to account for the many uncertainties. Thus......, a reliability assessment should be based on probabilistic methods where stochastic modeling of failures is performed. This thesis focuses on probabilistic models and the stochastic modeling of the fatigue life of the wind turbine drivetrain. Hence, two approaches are considered for stochastic modeling...

The analysis of pressurizer safety valve stuck open accident for low power and shutdown PSA

Energy Technology Data Exchange (ETDEWEB)

Lim, Ho Gon; Park, Jin Hee; Jang, Seong Chul; Kim, Tae Woon

2005-01-01

The PSV (Pressurizer Safety Valve) popping test carried out practically in the early phase of a refueling outage has a little possibility of triggering a test-induced LOCA due to a PSV not fully closed or stuck open. According to a KSNP (Korea Standard Nuclear Power Plant) low power and shutdown PSA (Probabilistic Safety Assessment), the failure of a HPSI (High Pressure Safety Injection) following a PSV stuck open was identified as a dominant accident sequence with a significant contribution to low power and shutdown risks. In this study, we aim to investigate the consequences of the NPP for the various accident sequences following the PSV stuck open as an initiating event through the thermal-hydraulic system code calculations. Also, we search the accident mitigation method for the sequence of HPSI failure, then, the applicability of the method is verified by the simulations using T/H system code.

Probabilistic assessment of leak-before-break

International Nuclear Information System (INIS)

Bush, S.H.

1984-01-01

A summary of results illustrating what might be derived from a probabilistic risk assessment (PRA) study follows. The failure probabilities for larger sizes of nuclear piping are considered to be in the range of 10 -4 to 10 -6 per reactor-year (exclusive of intergranular stress corrosion cracking (IGSCC). Smaller pipe sizes, of lesser safety significance, have much higher failure rates. In BWRs, IGSCC can cause failure rates much higher than 10 -4 in piping 4 to 10 in. in size. Suggested failure mechanisms apply in most instances, exclusive of IGSCC. Catastrophic failures would appear more likely from operator error or design and construction errors (water hammer, improper handling of dynamic loads, and undetected fabrication defects) rather than conventional flaw initiation and growth by fatigue

Internal Flooding Probabilistic Safety Assessment of an OPR-1000 Plant during Low Power and Shutdown Operation

International Nuclear Information System (INIS)

Lee, Yoon Hwan; Park, Jin Hee; Lim, Ho Gon

2016-01-01

In 2009, the electric power research institute (EPRI) published a guideline for the development of IF-PRA that addresses the requirements of the ASME/ANS RASa-2009 PRA consensus standard. The EPRI guideline delineates a level of detail and assessment complexity that has been significantly increased with respect to the guidance for IF assessment performed for the individual plant examination (IPE) to address Generic Letter 88- 20. The main differences include: A more systematic approach to the definition of flood area. The identification, screening and analysis of flooding sources and scenarios. The calculation of the initiating-event frequency (IEF) based on the actual length and characteristics of the piping. The inclusion of spatial effects such as spray from pipe leaks. The specific documentation associated with the plant walkdowns. Among these differences, this research focused on the third and fourth items when performing the internal flooding PSA. This is done by identifying the pipe and fluid characteristics, assessing the pipe pressure, characterizing the pipe (i.e., pipe diameter, length, etc.) and determining the pressure boundary failure frequency. The results were summed for the various piping systems within a given flood area to arrive at an overall internal flood initiating frequency for a given flood mode (i.e., spray, general flood, or major flood) for that particular area by each POS (Plant Operational State). In this initiating event frequency evaluations, the POS duration time is especially considered to get the real values for LPSD state. Characterizations of spray scenarios were evaluated to determine their impact on plant risk caused by internal flooding events.

PRA and the implementation of quantitative safety goals

International Nuclear Information System (INIS)

Okrent, D.

1983-01-01

With the adoption by the U.S. Nuclear Regulatory Commission (NRC) in January, 1983, of a Policy Statement on Safety Goals for the Operation of Nuclear Power Plants, probabilitstic risk assessment (PRA) has taken on increased importance in nuclear reactor safety. Although the Reactor Safety Study, WASH-1400, was a major pioneering effort that revolutionized thinking about reactor safety, PRA was used only on occasion by the NRC regulatory staff prior to the accident at Three Mile Island. Since then, PRA has been used more and more as an important factor in decision making, usually for specific issues. The nuclear industry has also employed PRA, sometimes to make its case on specific issues, sometimes to present a position on overall risk. The advent of the Zion and Indian Point PRAs, with their treatment of risks from fire, wind, and earthquakes, and their examination of the course of core melt accidents, has added a new dimension to the overall picture. Although the NRC has stated that during the next two year evolution period, its quantitative design objectives and PRA are not to enter directly into the licensing process, many important issues will be influenced significantly by the results of risk and reliability studies. In fact, PRA may be coming into a position of great importance before the methodology, data, and process are sufficiently mature for the task. Large gaps still exist in our understanding of phenomena and in input information; and much of the final result depends on subjective input; large differences of opinion can and should be expected to persist. Accepted standards for quality assurance, and adequacy and depth of independent, peer review remain to be formulated and achieved. This paper will summarize the recently adopted NRC safety policy and the two-year evaluation plan, and will provide, by example, some words of caution concerning a few of the difficulties which may arise. (orig.)

Evaluation of Frequency and Restoration time for Loss of Offsite Power events based on domestic operation experience

International Nuclear Information System (INIS)

Park, Jin-Hee; Han, Sang-Hoon; Lee, Ho Joong

2006-01-01

It is recognized that the availability of AC power to nuclear power plants is essential for safe operation and shutdown and accident recovery of commercial nuclear power plants (NPPs). Unavailability of AC power can be a important adverse impact on a plant's ability to recover accident sequences and maintain safe shutdown. The probabilistic safety assessment (PSA or PRA) performed for Korea NPPs also indicated that a loss of offsite power (LOOP) event and a station blackout (SBO) event can be a important contributors to total risk at nuclear power plant, contributing from 30% to 70% of the total risk at some NPPs in Korea. But, up to now, the LOOP and subsequent restoration time are important inputs to plant probabilistic risk assessment have relied upon foreign data. Therefore, in this paper, the actual LOOP events that have occurred from 1978 to 2004 at commercial nuclear power plants in Korea are collected. A statistical analysis for LOOP frequency and restoration time are performed to apply NPPs's specific and realistic risk model in Korea. Additionally, an engineering analysis is also performed to obtain the insights about the LOOP events

Failure Modes Taxonomy for Reliability Assessment of Digital Instrumentation and Control Systems for Probabilistic Risk Analysis - Failure modes taxonomy for reliability assessment of digital I and C systems for PRA

International Nuclear Information System (INIS)

Amri, A.; Blundell, N.; ); Authen, S.; Betancourt, L.; Coyne, K.; Halverson, D.; Li, M.; Taylor, G.; Bjoerkman, K.; Brinkman, H.; Postma, W.; Bruneliere, H.; Chirila, M.; Gheorge, R.; Chu, L.; Yue, M.; Delache, J.; Georgescu, G.; Deleuze, G.; Quatrain, R.; Thuy, N.; Holmberg, J.-E.; Kim, M.C.; Kondo, K.; Mancini, F.; Piljugin, E.; Stiller, J.; Sedlak, J.; Smidts, C.; Sopira, V.

2015-01-01

Digital protection and control systems appear as upgrades in older nuclear power plants (NPP), and are commonplace in new NPPs. To assess the risk of NPP operation and to determine the risk impact of digital systems, there is a need to quantitatively assess the reliability of the digital systems in a justifiable manner. Due to the many unique attributes of digital systems (e.g., functions are implemented by software, units of the system interact in a communication network, faults can be identified and handled online), a number of modelling and data collection challenges exist, and international consensus on the reliability modelling has not yet been reached. The objective of the task group called DIGREL has been to develop a taxonomy of failure modes of digital components for the purposes of probabilistic risk analysis (PRA). An activity focused on the development of a common taxonomy of failure modes is seen as an important step towards standardised digital instrumentation and control (I and C) reliability assessment techniques for PRA. Needs from PRA has guided the work, meaning, e.g., that the I and C system and its failures are studied from the point of view of their functional significance point of view. The taxonomy will be the basis of future modelling and quantification efforts. It will also help to define a structure for data collection and to review PRA studies. The proposed failure modes taxonomy has been developed by first collecting examples of taxonomies provided by the task group organisations. This material showed some variety in the handling of I and C hardware failure modes, depending on the context where the failure modes have been defined. Regarding the software part of I and C, failure modes defined in NPP PRAs have been simple - typically a software CCF failing identical processing units. The DIGREL task group has defined a new failure modes taxonomy based on a hierarchical definition of five levels of abstraction: 1. system level (complete

MATILDA Version-2: Rough Earth TIALD Model for Laser Probabilistic Risk Assessment in Hilly Terrain - Part II

Science.gov (United States)

2017-07-28

risk assessment for “unsafe” scenarios. Recently, attention in the DoD has turned to Probabilistic Risk Assessment (PRA) models [5,6] as an...corresponding to the CRA undershoot boundary. The magenta- coloured line represents the portion of the C-RX(U) circle that would contribute to the...Tertiary Precaution Surface. Undershoot related laser firing restrictions within the green- coloured C-RX(U) can be ignored. Figure 34

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

International Nuclear Information System (INIS)

Rahman, S.; Ghadiali, N.; Wilkowski, G.

1997-01-01

During the NRC's Short Cracks in Piping and Piping Welds Program at Battelle, a probabilistic methodology was developed to conduct fracture evaluations of circumferentially cracked pipes for application to leak-rate detection. Later, in the IPIRG-2 program, several parameters that may affect leak-before-break and other pipe flaw evaluations were identified. This paper presents new results from several uncertainty analyses to evaluate the effects of normal operating stresses, normal plus safe-shutdown earthquake stresses, off-centered cracks, restraint of pressure-induced bending, and dynamic and cyclic loading rates on the conditional failure probability of pipes. systems in BWR and PWR. For each parameter, the sensitivity to conditional probability of failure and hence, its importance on probabilistic leak-before-break evaluations were determined

Uncertainty analysis for probabilistic pipe fracture evaluations in LBB applications

Energy Technology Data Exchange (ETDEWEB)

Rahman, S.; Ghadiali, N.; Wilkowski, G.

1997-04-01

During the NRC`s Short Cracks in Piping and Piping Welds Program at Battelle, a probabilistic methodology was developed to conduct fracture evaluations of circumferentially cracked pipes for application to leak-rate detection. Later, in the IPIRG-2 program, several parameters that may affect leak-before-break and other pipe flaw evaluations were identified. This paper presents new results from several uncertainty analyses to evaluate the effects of normal operating stresses, normal plus safe-shutdown earthquake stresses, off-centered cracks, restraint of pressure-induced bending, and dynamic and cyclic loading rates on the conditional failure probability of pipes. systems in BWR and PWR. For each parameter, the sensitivity to conditional probability of failure and hence, its importance on probabilistic leak-before-break evaluations were determined.

Identification of human-induced initiating events in the low power and shutdown operation using the commission error search and assessment method

Energy Technology Data Exchange (ETDEWEB)

Kim, Yong Chan; Kim, Jong Hyun [KEPCO International Nuclear Graduate School (KINGS), Ulsan (Korea, Republic of)

2015-03-15

Human-induced initiating events, also called Category B actions in human reliability analysis, are operator actions that may lead directly to initiating events. Most conventional probabilistic safety analyses typically assume that the frequency of initiating events also includes the probability of human-induced initiating events. However, some regulatory documents require Category B actions to be specifically analyzed and quantified in probabilistic safety analysis. An explicit modeling of Category B actions could also potentially lead to important insights into human performance in terms of safety. However, there is no standard procedure to identify Category B actions. This paper describes a systematic procedure to identify Category B actions for low power and shutdown conditions. The procedure includes several steps to determine operator actions that may lead to initiating events in the low power and shutdown stages. These steps are the selection of initiating events, the selection of systems or components, the screening of unlikely operating actions, and the quantification of initiating events. The procedure also provides the detailed instruction for each step, such as operator's action, information required, screening rules, and the outputs. Finally, the applicability of the suggested approach is also investigated by application to a plant example.

The EBR-II Probabilistic Risk Assessment: Results and insights

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1993-01-01

This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1. 6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The probability of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquake) is 3.6 10 -6 yr -1 . overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double, vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

Handbook of methods for risk-based analysis of technical specifications

International Nuclear Information System (INIS)

Samanta, P.K.; Kim, I.S.; Mankamo, T.; Vesely, W.E.

1996-01-01

Technical Specifications (TS) requirements for nuclear power plants define the Limiting Conditions for Operations (LCOs) and Surveillance Requirements (SRs) to assure safety during operation. In general, these requirements are based on deterministic analyses and engineering judgments. Improvements in these requirements are facilitated by the availability of plant-specific Probabilistic Risk Assessments (PRAs). The US Nuclear Regulatory Commission (USNRC) Office of Research sponsored research to develop systematic, risk-based methods to improve various aspects of TS requirements. A handbook of methods summarizing such risk-based approaches has been completed in 1994. It is expected that this handbook will provide valuable input to NRC's present work in developing guidance for using PRA in risk-informed regulation. The handbook addresses reliability and risk-based methods for evaluating allowed outage times (AOTs), action statements requiring shutdown where shutdown risk may be substantial, surveillance test intervals (STIs), managing plant configurations, and scheduling maintenance

PRA (probabilistic risk analysis) in the nuclear sector. Quantifying human error and human malice

International Nuclear Information System (INIS)

Heyes, A.G.

1995-01-01

Regardless of the regulatory style chosen ('command and control' or 'functional') a vital prerequisite for coherent safety regulations in the nuclear power industry is the ability to assess accident risk. In this paper we present a critical analysis of current techniques of probabilistic risk analysis applied in the industry, with particular regard to the problems of quantifying risks arising from, or exacerbated by, human risk and/or human error. (Author)

Procedures for the elicitation of expert judgements in the probabilistic risk analysis of the long-term effects of radioactive waste repositories: an annotated bibliography

International Nuclear Information System (INIS)

Watson, S.R.

1993-01-01

This annotated bibliography describes the key literature relevant to the elicitation of expert judgements in radioactive waste management. The bibliography is divided into seven sections; section 2 lists the literature exploring the proper interpretation of probabilities used in Probabilistic Risk Analysis (PRA). Section 3 lists literature describing other calculi for handling uncertainty in a numerical fashion. In section 4 comments are given on how to elicit probabilities from individuals as a measure of subjective degrees of belief and section 5 lists the literature concerning how expert judgements can be combined. Sections 6 and 7 list literature giving an overview of the issues involved in PRA for radioactive waste repositories. (author)

Applicability of PRISM PRA Methodology to the Level II Probabilistic Safety Analysis of KALIMER-600 (I) (Core Damage Event Tree Analysis Part)

International Nuclear Information System (INIS)

Park, S. Y.; Kim, T. W.; Ha, K. S.; Lee, B. Y.

2009-03-01

The Korea Atomic Energy Research Institute (KAERI) has been developing liquid metal reactor (LMR) design technologies under a National Nuclear R and D Program. Nevertheless, there is no experience of the PSA domestically for a fast reactor with the metal fuel. Therefore, the objective of this study is to establish the methodologies of risk assessment for the reference design of KALIMER-600 reactor. An applicability of the PSA of the PRISM plant to the KALIMER-600 has been studied. The study is confined to a core damage event tree analysis which is a part of a level 2 PSA. Assuming that the accident types, which can be developed from level 1 PSA, are same as the PRISM PRA, core damage categories are defined and core damage event trees are developed for the KALIMER-600 reactor. Fission product release fractions of the core damage categories and branch probabilities of the core damage event trees are referred from the PRISM PRA temporarily. Plant specific data will be used during the detail analysis

NRC Support for the Kalinin (VVER) probabilistic risk assessment

International Nuclear Information System (INIS)

Bley, D.; Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T.; Johnson, D.; Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M.

1998-01-01

The US Nuclear Regulatory Commission (NRC) and the Federal Nuclear and Radiation Safety Authority of the Russian Federation have been working together since 1994 to carry out a probabilistic risk assessment (PRA) of a VVER-1000 in the Russian Federation. This was a recognition by both parties that this technology has had a profound effect on the discipline of nuclear reactor safety in the West and that the technology should be transferred to others so that it can be applied to Soviet-designed plants. The NRC provided funds from the Agency for International Development and technical support primarily through Brookhaven National Laboratory and its subcontractors. The latter support was carried out through workshops, by documenting the methodology to be used in a set of guides, and through periodic review of the technical activity. The result of this effort to date includes a set of procedure guides, a draft final report on the Level 1 PRA for internal events (excluding internal fires and floods), and progress reports on the fire, flood, and seismic analysis. It is the authors belief that the type of assistance provided by the NRC has been instrumental in assuring a quality product and transferring important technology for use by regulators and operators of Soviet-designed reactors. After a thorough review, the report will be finalized, lessons learned will be applied in the regulatory and operational regimes in the Russian Federation, and consideration will be given to supporting a containment analysis in order to complete a simplified Level 2 PRA

Diablo Canyon internal events PRA [Probabilistic Risk Assessment] review: Methodology and findings

International Nuclear Information System (INIS)

Fitzpatrick, R.G.; Bozoki, G.; Sabek, M.

1990-01-01

The review of the Diablo Canyon Probabilistic Risk Assessment (DCRPA) incorporated some new and innovative approaches. These were necessitated by the unprecedented size, scope and level of detail of the DCRPA, which was submitted to the NRC for licensing purposes. This paper outlines the elements of the internal events portion of the review citing selected findings to illustrate the various approaches employed. The paper also provides a description of the extensive and comprehensive importance analysis applied by BNL to the DCRPA model. Importance calculations included: top event/function level; individual split fractions; pair importances between frontline-support and support-support systems; system importance by initiator; and others. The paper concludes with a brief discussion of the effectiveness of the applied methodology. 3 refs., 5 tabs

Probabilistic risk analysis for nuclear power plants

International Nuclear Information System (INIS)

Hauptmanns, U.

1988-01-01

Risk analysis is applied if the calculation of risk from observed failures is not possible, because events contributing substantially to risk are too seldom, as in the case of nuclear reactors. The process of analysis provides a number of benefits. Some of them are listed. After this by no means complete enumeration of possible benefits to be derived from a risk analysis. An outline of risk studiesd for PWR's with some comments on the models used are given. The presentation is indebted to the detailed treatment of the subject given in the PRA Procedures Guide. Thereafter some results of the German Risk Study, Phase B, which is under way are communicated. The paper concludes with some remarks on probabilistic considerations in licensing procedures. (orig./DG)

A risk assessment of the SAFR plant

International Nuclear Information System (INIS)

Rutherford, P.D.; Mills, J.C.; Lancet, R.T.; Nourjah, P.

1987-01-01

The Sodium Advanced Fast Reactor (SAFR) is a modular, advanced concept, Liquid Metal Reactor (LMR), funded by the U.S., and designed by Rockwell International, Bechtel Corporation, and Combustion Engineering. SAFR utilizes the inherently safe features of small fast reactors, including natural convection decay heat removal systems, a self-actuated shutdown system (SASS) and inherent core response to design basis events without scram including transient overpower (TOP), loss of flow (LOF), and loss of heat sink (LOHS) events. A Level 3 probabilistic risk assessment (PRA) has been performed which demonstrates considerable reduction in plant and public risk compared to current commercial reactors. (orig./HSCH)

A probabilistic evaluation of temporarily extending the diesel generator limiting condition for operation

International Nuclear Information System (INIS)

Kukielka, C.A.; Detamore, M.B.

1986-01-01

Pennsylvania Power and Light Company is installing a fifth diesel generator at its Susquehanna Steam Electric Station. This new diesel generator will serve as a ''swing'' diesel generator, i.e., perform the function of any of the existing diesel generators. The addition of this diesel generator avoids a dual unit shutdown should a diesel generator outage for maintenance or repair exceed the three day Technical Specification Limiting Condition for Operation (LCO). It is estimated that up to fifteen days per diesel generator will be required to complete the tie-in. In accordance with the existing three day LCO, a dual unit shutdown would then result. To preclude such a shutdown, a one time temporary extension to the three day LCO was requested of the Nuclear Regulatory Commission (NRC). This paper presents a probabilistic evaluation that was used to estimate the increased risk due to extending the LCO to a cumulative sixty days to allow the tie-in of the fifth diesel generator

Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents

International Nuclear Information System (INIS)

Houghton, W.J.

1980-06-01

A probabilistic risk assessment (PRA) approach is proposed to be used to scrutinize selection of accident sequences. A technique is described in this Licensing Topical Report to identify candidates for Design Basis Accidents (DBAs) utilizing the risk assessment results. As a part of this technique, it is proposed that events with frequencies below a specified limit would not be candidates. The use of the methodology described is supplementary to the traditional, deterministic approach and may result, in some cases, in the selection of multiple failure sequences as DBAs; it may also provide a basis for not considering some traditionally postulated events as being DBAs. A process is then described for selecting a list of DBAs based on the candidates from PRA as supplementary to knowledge and judgments from past licensing practice. These DBAs would be the events considered in Chapter 15 of Safety Analysis Reports of high-temperature gas-cooled reactors

Load out and offshore lifting of the PRA-1 platform modules; Embarque e icamento offshore dos modulos de PRA-1

Energy Technology Data Exchange (ETDEWEB)

Barbosa, Fernando; Raigorodsky, Jacques; Mitidieri, Jorge L.U.; Ricardi, Paulo S. [Construtora Norberto Odebrecht S.A., Rio de Janeiro, RJ (Brazil)

2008-07-01

The technology innovations are characteristics of offshore Engineering around the world. These technologies just make sense when they aim the productivity, security and costs gains compared to ordinary methods. It is in this context that the proposal of the Consorcio PRA-1 (Odebrecht e UTC) team makes sense, in the definition of basic methodology for the PRA-1 platform construction and installation. Through the innovative concept, It was defined (still in the proposal phase) the basic premise that the modules construction and assembly were onshore ending up that just few hours after the offshore installation the modules should be operational in minimal habitability conditions. This innovative method allowed the lack of Flotel, that is a platform which provide support to the offshore construction and assembly (Flotel represents a high costs to the project) and, as consequence, the contract signature by CONSORCIO PRA-1. This work aims to describe the method used for the LOUD-OUT of the PRA-1 modules and the installation of them on the jacket through a vessel provide with cranes the has performed the lifting. Theses operations became unique in Brazil due its challengers characteristics: Module 12 weight = 7203 tf and Module 35 = 5725 tf. For the accomplishment of the Load-out and offshore lifting, was performed a detailed planning and a high level of subcontract interface management. The operations mentioned above were filmed/photographed and published in the specialized media. (author)

Standardized approach for developing probabilistic exposure factor distributions

Energy Technology Data Exchange (ETDEWEB)

Maddalena, Randy L.; McKone, Thomas E.; Sohn, Michael D.

2003-03-01

The effectiveness of a probabilistic risk assessment (PRA) depends critically on the quality of input information that is available to the risk assessor and specifically on the probabilistic exposure factor distributions that are developed and used in the exposure and risk models. Deriving probabilistic distributions for model inputs can be time consuming and subjective. The absence of a standard approach for developing these distributions can result in PRAs that are inconsistent and difficult to review by regulatory agencies. We present an approach that reduces subjectivity in the distribution development process without limiting the flexibility needed to prepare relevant PRAs. The approach requires two steps. First, we analyze data pooled at a population scale to (1) identify the most robust demographic variables within the population for a given exposure factor, (2) partition the population data into subsets based on these variables, and (3) construct archetypal distributions for each subpopulation. Second, we sample from these archetypal distributions according to site- or scenario-specific conditions to simulate exposure factor values and use these values to construct the scenario-specific input distribution. It is envisaged that the archetypal distributions from step 1 will be generally applicable so risk assessors will not have to repeatedly collect and analyze raw data for each new assessment. We demonstrate the approach for two commonly used exposure factors--body weight (BW) and exposure duration (ED)--using data for the U.S. population. For these factors we provide a first set of subpopulation based archetypal distributions along with methodology for using these distributions to construct relevant scenario-specific probabilistic exposure factor distributions.

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

Energy Technology Data Exchange (ETDEWEB)

Yamano, Hidemasa, E-mail: yamano.hidemasa@jaea.go.jp; Nishino, Hiroyuki; Kurisaka, Kenichi

2016-11-15

Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10{sup −6}/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

Development of probabilistic risk assessment methodology against extreme snow for sodium-cooled fast reactor

International Nuclear Information System (INIS)

Yamano, Hidemasa; Nishino, Hiroyuki; Kurisaka, Kenichi

2016-01-01

Highlights: • Snow PRA methodology was developed. • Snow hazard category was defined as the combination of daily snowfall depth (speed) and snowfall duration. • Failure probability models of snow removal action, manual operation of the air cooler dampers and the access route were developed. • Snow PRA showed less than 10"−"6/reactor-year of core damage frequency. - Abstract: This paper describes snow probabilistic risk assessment (PRA) methodology development through external hazard and event sequence evaluations mainly in terms of decay heat removal (DHR) function of a sodium-cooled fast reactor (SFR). Using recent 50-year weather data at a typical Japanese SFR site, snow hazard categories were set for the combination of daily snowfall depth (snowfall speed) and snowfall duration which can be calculated by dividing the snow depth by the snowfall speed. For each snow hazard category, the event sequence was evaluated by event trees which consist of several headings representing the loss of DHR. Snow removal action and manual operation of the air cooler dampers were introduced into the event trees as accident managements. Access route failure probability model was also developed for the quantification of the event tree. In this paper, the snow PRA showed less than 10"−"6/reactor-year of core damage frequency. The dominant snow hazard category was the combination of 1–2 m/day of snowfall speed and 0.5–0.75 day of snowfall duration. Importance and sensitivity analyses indicated a high risk contribution of the securing of the access routes.

Level 1 probabilistic risk assessment of low power and shutdown operations at a PWR: Phase 2 results

International Nuclear Information System (INIS)

Chu, T.L.; Bozoki, G.; Kohut, P.; Musicki, Z.; Wong, S.M.; Yang, J.; Hsu, C.J.; Diamond, D.J.; Su, R.F.; Holmes, B.; Siu, N.; Bley, D.; Lin, J.

1992-01-01

As a result of the Chernobyl accident and other precursor events (e.g., Diablo Canyon), the US Nuclear Regulatory Commission's (NRC's) Office of Nuclear Regulatory Research (RES) initiated an extensive project during 1989 to carefully examine the potential risks during Low Power and Shutdown (LP ampersand S) operations. Shortly after the program began, an event occurred at the Vogtle plant during shutdown, which further intensified the effort of the LP ampersand S program. In the LP ampersand S program, one pressurized water reactor (PWR), Surry, and one boiling water reactor (BWR), Grand Gulf, were selected, mainly because they were previously analyzed in the NUREG-1150 Study. The Level-1 Program is being performed in two phases. Phase 1 was dedicated to performing a coarse screening level-1 analysis including internal fire and flood. A draft report was completed in November, 1991. In the phase 2 study, mid-loop operations at the Surry plant were analyzed in detail. The objective of this paper is to present the approach of the phase 2 study and the preliminary results and insights

Probabilistic Risk Assessment for Decision Making During Spacecraft Operations

Science.gov (United States)

Meshkat, Leila

2009-01-01

Decisions made during the operational phase of a space mission often have significant and immediate consequences. Without the explicit consideration of the risks involved and their representation in a solid model, it is very likely that these risks are not considered systematically in trade studies. Wrong decisions during the operational phase of a space mission can lead to immediate system failure whereas correct decisions can help recover the system even from faulty conditions. A problem of special interest is the determination of the system fault protection strategies upon the occurrence of faults within the system. Decisions regarding the fault protection strategy also heavily rely on a correct understanding of the state of the system and an integrated risk model that represents the various possible scenarios and their respective likelihoods. Probabilistic Risk Assessment (PRA) modeling is applicable to the full lifecycle of a space mission project, from concept development to preliminary design, detailed design, development and operations. The benefits and utilities of the model, however, depend on the phase of the mission for which it is used. This is because of the difference in the key strategic decisions that support each mission phase. The focus of this paper is on describing the particular methods used for PRA modeling during the operational phase of a spacecraft by gleaning insight from recently conducted case studies on two operational Mars orbiters. During operations, the key decisions relate to the commands sent to the spacecraft for any kind of diagnostics, anomaly resolution, trajectory changes, or planning. Often, faults and failures occur in the parts of the spacecraft but are contained or mitigated before they can cause serious damage. The failure behavior of the system during operations provides valuable data for updating and adjusting the related PRA models that are built primarily based on historical failure data. The PRA models, in turn

Reactor shutdown device

International Nuclear Information System (INIS)

Inoue, Toyokazu.

1982-01-01

Purpose: To obtain a highly reliable reactor shutdown device capable of checking its function irrespective of the state whether shutdown or operation in a gas-cooled type reactor. Constitution: A hopper is disposed above a guide tube inserted into the reactor core and particulate neutron absorbers are contained in the hopper. An opening for falling particles is disposed to the bottom of the hopper in opposition to the upper end of the guide pipe and the opening is closed by a plug suspended by way of a weld line so as to be capable of dropping. A power source for supplying electrical current to the weld line is disposed. Accordingly, if the current is supplied to the weld line, the line is cut by welding to fall the plug so that the neutron-absorbing particles fall from the opening into the guide pipe to shutdown the reactor, whereby high reliability is obtained for the operation. (Seki, T.)

A socio-technical, probabilistic risk assessment model for surgical site infections in ambulatory surgery centers.

Science.gov (United States)

Bish, Ebru K; El-Amine, Hadi; Steighner, Laura A; Slonim, Anthony D

2014-10-01

To understand how structural and process elements may affect the risk for surgical site infections (SSIs) in the ambulatory surgery center (ASC) environment, the researchers employed a tool known as socio-technical probabilistic risk assessment (ST-PRA). ST-PRA is particularly helpful for estimating risks in outcomes that are very rare, such as the risk of SSI in ASCs. Study objectives were to (1) identify the risk factors associated with SSIs resulting from procedures performed at ASCs and (2) design an intervention to mitigate the likelihood of SSIs for the most common risk factors that were identified by the ST-PRA for a particular surgical procedure. ST-PRA was used to study the SSI risk in the ASC setting. Both quantitative and qualitative data sources were utilized, and sensitivity analysis was performed to ensure the robustness of the results. The event entitled "fail to protect the patient effectively" accounted for 51.9% of SSIs in the ambulatory care setting. Critical components of this event included several failure risk points related to skin preparation, antibiotic administration, staff training, proper response to glove punctures during surgery, and adherence to surgical preparation rules related to the wearing of jewelry, watches, and artificial nails. Assuming a 75% reduction in noncompliance on any combination of 2 of these 5 components, the risk for an SSI decreased from 0.0044 to between 0.0027 and 0.0035. An intervention that targeted the 5 major components of the major risk point was proposed, and its implications were discussed.

Probabilistic risk assessment using event tables and the BNL [Brookhaven National Laboratory] event-tree analyzer

International Nuclear Information System (INIS)

Fullwood, R.R.; Shier, W.G.

1989-01-01

Probabilistic risk analysis (PRA) is being used to study design alternatives for the advanced neutron source research reactor being designed at Oak Ridge National Laboratory for operation in the 1990s. Major communication paths between the designers and the safety analysts are accident discussions supported by event tables, event-tree graphics, and accident sequence probabilities. The BETA code used in conjunction with a word processor provides this linkage. This paper describes the process, features of the BETA, how it works, and some examples of usage

Development of risk assessment methodology against natural external hazards for sodium-cooled fast reactors: project overview and strong Wind PRA methodology - 15031

International Nuclear Information System (INIS)

Yamano, H.; Nishino, H.; Kurisaka, K.; Okano, Y.; Sakai, T.; Yamamoto, T.; Ishizuka, Y.; Geshi, N.; Furukawa, R.; Nanayama, F.; Takata, T.; Azuma, E.

2015-01-01

This paper describes mainly strong wind probabilistic risk assessment (PRA) methodology development in addition to the project overview. In this project, to date, the PRA methodologies against snow, tornado and strong wind were developed as well as the hazard evaluation methodologies. For the volcanic eruption hazard, ash fallout simulation was carried out to contribute to the development of the hazard evaluation methodology. For the forest fire hazard, the concept of the hazard evaluation methodology was developed based on fire simulation. Event sequence assessment methodology was also developed based on plant dynamics analysis coupled with continuous Markov chain Monte Carlo method in order to apply to the event sequence against snow. In developing the strong wind PRA methodology, hazard curves were estimated by using Weibull and Gumbel distributions based on weather data recorded in Japan. The obtained hazard curves were divided into five discrete categories for event tree quantification. Next, failure probabilities for decay heat removal related components were calculated as a product of two probabilities: i.e., a probability for the missiles to enter the intake or out-take in the decay heat removal system, and fragility caused by the missile impacts. Finally, based on the event tree, the core damage frequency was estimated about 6*10 -9 /year by multiplying the discrete hazard probabilities in the Gumbel distribution by the conditional decay heat removal failure probabilities. A dominant sequence was led by the assumption that the operators could not extinguish fuel tank fire caused by the missile impacts and the fire induced loss of the decay heat removal system. (authors)

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

International Nuclear Information System (INIS)

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

1994-06-01

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

A desktop PRA

International Nuclear Information System (INIS)

Dolan, B.J.; Weber, B.J.

1989-01-01

This paper reports that Duke Power Company has completed full-scope PRAs for each of its nuclear stations - Oconee, McGuire and Catawba. These living PRAs are being maintained using desktop personal computers. Duke's PRA group now has powerful personal computer-based tools that have both decreased direct costs (computer analysis expenses) and increased group efficiency (less time to perform analyses). The shorter turnaround time has already resulted in direct savings through analyses provided in support of justification for continued station operation. Such savings are expected to continue with similar future support

Shutdown risk monitoring in TEPCO

International Nuclear Information System (INIS)

Sato, Hiroki; Masuda, Takahiro; Denda, Yasutaka; Yoneyama, Mitsuru; Imai, Shun-ichi; Miyata, Koichi

2009-01-01

At present, we are introducing risk monitors into our all three nuclear power stations; Fukushima Daiichi, Fukushima Daini and Kashiwazaki Kariwa, with technical support of TEPSYS. By monitoring shutdown risk of each unit, we are trying to optimize risks during outage inspection, and raising staff's awareness for reactor safety. This paper presents our recent shutdown risk monitoring activities in Fukushima Daiichi NPS. Shutdown risk monitoring has been carried out for the past five outages of Fukushima Daiichi NPS. Daily-changing shutdown risk is evaluated in the form of core damage frequency (CDF [/day/reactor]). We also examine high-risk point of outage plan if CDF is greater than the threshold at anytime of outage. The results are delivered to operational and maintenance staff before outage. The threshold value is set ten times as much as CDF of unit in operation. As CDF exceeds the threshold, we try to either change the system configuration, or let workers pay more attention to their works during the high-risk period. We already have some examples of outage plan modification to reduce CDF using the risk monitoring information. Greater number of station staff tends to pay more attention to shutdown risk thanks to these activities. (author)

The NUREG-1150 probabilistic risk assessment for the Sequoyah nuclear plant

International Nuclear Information System (INIS)

Gregory, J.J.; Breeding, R.J.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.; Murfin, W.B.

1992-01-01

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Sequoyah Nuclear Plant performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, the accident progression, source term, and consequence analyses, and the risk results obtained when the results of these analyses are combined with the accident frequency analysis. The results of this PRA indicate that the offsite risk from internal initiating events at Sequoyah are quite low with respect to the safety goals. The containment appears likely to withstand the loads that might be placed upon it if the reactor vessel fails. A good portion of the risk, in this analysis, comes from initiating events which bypass the containment. These events are estimated to have a relatively low frequency of occurrence, but their consequences are quite large. Other events that contribute to offsite risk involve early containment failures that occur during degradation of the core or near the time of vessel breach. Considerable uncertainty is associated with the risk estimates produced in this analysis. Offsite risk from external initiating events was not included in this analysis. (orig.)

Magnetic disconnect for secondary shutdown

International Nuclear Information System (INIS)

Lessor, D.L.

1972-01-01

A description is given of studies to develop a magnetic holding clutch in the control rod drive line as an alternate shutdown device for the FFTF. Results indicate that a three-phase disconnect, hold, and backup shutdown system can be designed to operate satisfactorily. (U.S.)

Risks due to fires at Big Rock Point

International Nuclear Information System (INIS)

Brinsfield, W.A.; Blanchard, D.P.

1983-01-01

The unique and older designs of the Big Rock Point nuclear plant is such that fires contribute significantly to the probability of core damage predicted in the probabilistic risk assessment performed for this plant. The methodology employed to determine this contribution reflects the unique, as constructed, plant design, while systematically and logically addressing the true effect of fires on the operation of the plant and the safety of the public. As a result of the methodology utilized in the PRA, recommendations are made which minimize the risk of core damage due to fires. Included in these recommendations is a proposal for equipment and controls to be included on the Big Rock Point alternate shutdown panel

Seismic Probabilistic Risk Assessment (SPRA), approach and results

International Nuclear Information System (INIS)

Campbell, R.D.

1995-01-01

During the past 15 years there have been over 30 Seismic Probabilistic Risk Assessments (SPRAs) and Seismic Probabilistic Safety Assessments (SPSAs) conducted of Western Nuclear Power Plants, principally of US design. In this paper PRA and PSA are used interchangeably as the overall process is essentially the same. Some similar assessments have been done for reactors in Taiwan, Korea, Japan, Switzerland and Slovenia. These plants were also principally US supplied or built under US license. Since the restructuring of the governments in former Soviet Bloc countries, there has been grave concern regarding the safety of the reactors in these countries. To date there has been considerable activity in conducting partial seismic upgrades but the overall quantification of risk has not been pursued to the depth that it has in Western countries. This paper summarizes the methodology for Seismic PRA/PSA and compares results of two partially completed and two completed PRAs of soviet designed reactors to results from earlier PRAs on US Reactors. A WWER 440 and a WWER 1000 located in low seismic activity regions have completed PRAs and results show the seismic risk to be very low for both designs. For more active regions, partially completed PRAs of a WWER 440 and WWER 1000 located at the same site show the WWER 440 to have much greater seismic risk than the WWER 1000 plant. The seismic risk from the 1000 MW plant compares with the high end of seismic risk for earlier seismic PRAs in the US. Just as for most US plants, the seismic risk appears to be less than the risk from internal events if risk is measured is terms of mean core damage frequency. However, due to the lack of containment for the earlier WWER 440s, the risk to the public may be significantly greater due to the more probable scenario of an early release. The studies reported have not taken the accident sequences beyond the stage of core damage hence the public heath risk ratios are speculative. (author)

Probabilistic risk assessment of insecticide concentrations in agricultural surface waters: a critical appraisal.

Science.gov (United States)

Stehle, Sebastian; Knäbel, Anja; Schulz, Ralf

2013-08-01

Due to the specific modes of action and application patterns of agricultural insecticides, the insecticide exposure of agricultural surface waters is characterized by infrequent and short-term insecticide concentration peaks of high ecotoxicological relevance with implications for both monitoring and risk assessment. Here, we apply several fixed-interval strategies and an event-based sampling strategy to two generalized and two realistic insecticide exposure patterns for typical agricultural streams derived from FOCUS exposure modeling using Monte Carlo simulations. Sampling based on regular intervals was found to be inadequate for the detection of transient insecticide concentrations, whereas event-triggered sampling successfully detected all exposure incidences at substantially lower analytical costs. Our study proves that probabilistic risk assessment (PRA) concepts in their present forms are not appropriate for a thorough evaluation of insecticide exposure. Despite claims that the PRA approach uses all available data to assess exposure and enhances risk assessment realism, we demonstrate that this concept is severely biased by the amount of insecticide concentrations below detection limits and therefore by the sampling designs. Moreover, actual insecticide exposure is of almost no relevance for PRA threshold level exceedance frequencies and consequential risk assessment outcomes. Therefore, we propose a concept that features a field-relevant ecological risk analysis of agricultural insecticide surface water exposure. Our study quantifies for the first time the environmental and economic consequences of inappropriate monitoring and risk assessment concepts used for the evaluation of short-term peak surface water pollutants such as insecticides.

Hiperurisemia pada Pra Diabetes

Directory of Open Access Journals (Sweden)

Ellyza Nasrul

2012-09-01

Full Text Available AbstrakAsam urat (AU merupakan produk akhir dari katabolisme adenin dan guanin yang berasal dari pemecahannukleotida purin. Urat dihasilkan oleh sel yang mengandung xanthine oxidase, terutama hepar dan usus kecil.Hiperurisemia adalah keadaan kadar asam urat dalam darah lebih dari 7,0 mg/dL.Pra diabetes adalah subjek yangmempunyai kadar glukosa plasma meningkat akan tetapi peningkatannya masih belum mencapai nilai minimaluntuk kriteria diagnosis diabetes melitus (DM. Glukosa darah puasa terganggu merupakan keadaan dimanapeningkatan kadar FPG≥100 mg/dL dan <126 mg/dL. Toleransi glukosa terganggu merupakan peningkatanglukosa plasma 2 jam setelah pembebanan 75 gram glukosa oral (≥140 mg/dL dan <200mg/dL dengan FPG<126 mg/dL.Insulin juga berperan dalam meningkatkan reabsorpsi asam urat di tubuli proksimal ginjal. Sehinggapada keadaan hiperinsulinemia pada pra diabetes terjadi peningkatan reabsorpsi yang akan menyebabkanhiperurisemia. Transporter urat yang berada di membran apikal tubuli renal dikenal sebagai URAT-1 berperandalam reabsorpsi urat.Kata kunci: Hiperurisemia, Pra DiabetesAbstractUric acid (AU is the end product of the catabolism of adenine and guanine nucleotides derived from thebreakdown of purines. Veins produced by cells containing xanthine oxidase, especially the liver and small intestine.Hyperuricemia is a state in the blood uric acid levels over 7.0 mg / dL.Pre-diabetes is a subject which has a plasmaglucose level will rise but the increase is still not reached the minimum value for the diagnostic criteria for diabetesmellitus (DM. Impaired fasting blood glucose is a condition in which increased levels of FPG ≥ 100 mg / dL and<126 mg / dL. Impaired glucose tolerance is an increase in plasma glucose 2 hours after 75 gram oral glucose load(≥ 140 mg / dL and <200mg/dl with FPG <126 mg / dL.Insulin also plays a role in increasing the reabsorption ofuric acid in renal proximal tubule. So that the hyperinsulinemia in the pre

Proceedings of workshop on reactor shutdown system

International Nuclear Information System (INIS)

1997-03-01

India has gained considerable experience in design, development, construction and operation of research and power reactors during the last four decades. Reactor shutdown system (RSS) is the most important engineered safety system of any reactor. A lot of technological developments have taken place to improve the reactor shutdown systems, particularly with advancement in reliability analysis and instrumentation and control. If the reactor is not shutdown, the fuel may melt, releasing radioactivity and possibly reactivity addition as in the case of Fast Breeder Reactor (FBR). Apart from radiological safety consequences, large investment has to be written off. The function of the RSS is to stop fission chain reaction and prevent breach of fuel. The design of RSS is multidisciplinary. It requires reactor physics analysis, design of absorber rods, drive mechanisms, safety logic to order shutdown and instrumentation to detect unsafe conditions. High reliability is essential and this requires two independent shutdown systems. This book contains the proceedings of the workshop on reactor shutdown system and papers relevant to INIS are indexed separately

Use of probabilistic risk assessments to define areas of possible exemption from regulatory requirements

International Nuclear Information System (INIS)

Thompson, C.A.; Carlson, D.; Kolaczkowski, A.; LaChance, J.

1988-01-01

The Risk-Based Licensing Program (RBLP) was sponsored by the Department of Energy for the purpose of establishing and demonstrating an approach for identifying potential areas for exemption from current regulatory requirements in the licensing of nuclear power plants. Such an approach could assist in the improvement of the regulatory process for both current and future nuclear plant designs. Use of the methodology could result in streamlining the regulatory process by eliminating unnecessarily detailed reviews of portions of a plant design not important to risk. The RBLP methodology utilizes probabilistic risk assessments, (PRAs), which are required of all future applicants for nuclear power plant licenses. PRA results are used as a screening tool to determine the risk significance of various plant features which are correlated to the risk importance of regulations to identify potential areas for regulatory exemption. Additional consideration is then given to non-risk factors in the final determination of exemption candidates. The RBLP methodology was demonstrated using an existing PRA. The results of the demonstration are highlighted. 10 refs

Simulation of Darlington shutdown and regulation systems

International Nuclear Information System (INIS)

1986-10-01

This report describes the development of a simulation of the Darlington Nuclear Generating Station shutdown and regulating systems, DARSIM. The DARSIM program simulates the spatial neutron dynamics, the regulation of the reactor power, and Shutdown System 1, SDS1, and Shutdown System 2, SDS2, software. The DARSIM program operates in the interactive simulation (INSIM) program environment

Safety shutdown separators

Science.gov (United States)

Carlson, Steven Allen; Anakor, Ifenna Kingsley; Farrell, Greg Robert

2015-06-30

The present invention pertains to electrochemical cells which comprise (a) an anode; (b) a cathode; (c) a solid porous separator, such as a polyolefin, xerogel, or inorganic oxide separator; and (d) a nonaqueous electrolyte, wherein the separator comprises a porous membrane having a microporous coating comprising polymer particles which have not coalesced to form a continuous film. This microporous coating on the separator acts as a safety shutdown layer that rapidly increases the internal resistivity and shuts the cell down upon heating to an elevated temperature, such as 110.degree. C. Also provided are methods for increasing the safety of an electrochemical cell by utilizing such separators with a safety shutdown layer.

Probabilistic risk assessment for back-end facilities: Improving the treatment of fire and explosion scenarios

International Nuclear Information System (INIS)

Sunman, C.R.J.; Campbell, R.J.; Wakem, M.J.

1996-01-01

The nuclear reprocessing facilities at Sellafield are a key component of the International business of BNFL. The operations carried out at the site extend from the receipt and storage of irradiated fuel, chemical reprocessing, plutonium and uranium finishing, through mixed oxide fuel production. Additionally there are a wide range of supporting processes including solid waste encapsulation, vitrification, liquid waste evaporation and treatment. Decommissioning of the site's older facilities is also proceeding. The comprehensive range of these activities requires that the safety assessment team keeps up to date with developments in the field, as well as conducting and sponsoring appropriate research into methodologies and modelling in order to deliver a cost effective, timely service. This paper will review the role of Probabilistic Risk Assessment (PRA) in safety cases for operations at Sellafield and go on to describe some areas of PRA methodology development in the UK and in which BNFL is a contributor. Finally the paper will summarise some specific areas of methodology development associated with improving the modelling of fire and explosion hazards which are specific to BNFL. (author)

Development of probabilistic risk analysis library

International Nuclear Information System (INIS)

Soga, Shota; Kirimoto, Yukihiro; Kanda, Kenichi

2015-01-01

We developed a library that is designed to perform level 1 Probabilistic Risk Analysis using Binary Decision Diagram (BDD). In particular, our goal is to develop a library that will allow Japanese electric utilities to take the advantages of BDD that can solve Event Tree (ET) and Fault Tree (FT) models analytically. Using BDD, the library supports negation in FT which allows more flexible modeling of ET/FT. The library is written by C++ within an object-oriented framework using open source software. The library itself is a header-only library so that Japanese electric utilities can take advantages of its transparency to speed up development and to build their own software for their specific needs. In this report, the basic capabilities of the library is briefly described. In addition, several applications of the library are demonstrated including validation of MCS evaluation of PRA model and evaluation of corrective and preventive maintenance considering common cause failure. (author)

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

International Nuclear Information System (INIS)

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

1994-06-01

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

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

Energy Technology Data Exchange (ETDEWEB)

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

1994-06-01

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

ORNL Isotopes Facilities Shutdown Program Plan

International Nuclear Information System (INIS)

Gibson, S.M.; Patton, B.D.; Sears, M.B.

1990-10-01

This plan presents the results of a technical and economic assessment for shutdown of the Oak Ridge National Laboratory (ORNL) isotopes production and distribution facilities. On December 11, 1989, the Department of Energy (DOE), Headquarters, in a memorandum addressed to DOE Oak Ridge Operations Office (DOE-ORO), gave instructions to prepare the ORNL isotopes production and distribution facilities, with the exception of immediate facility needs for krypton-85, tritium, and yttrium-90, for safe shutdown. In response to the memorandum, ORNL identified 17 facilities for shutdown. Each of these facilities is located within the ORNL complex with the exception of Building 9204-3, which is located at the Y-12 Weapons Production Plant. These facilities have been used extensively for the production of radioactive materials by the DOE Isotopes Program. They currently house a large inventory of radioactive materials. Over the years, these aging facilities have inherited the problems associated with storing and processing highly radioactive materials (i.e., facilities' materials degradation and contamination). During FY 1990, ORNL is addressing the requirements for placing these facilities into safe shutdown while maintaining the facilities under the existing maintenance and surveillance plan. The day-to-day operations associated with the surveillance and maintenance of a facility include building checks to ensure that building parameters are meeting the required operational safety requirements, performance of contamination control measures, and preventative maintenance on the facility and facility equipment. Shutdown implementation will begin in FY 1993, and shutdown completion will occur by the end of FY 1994

Design philosophy of PFBR shutdown systems

International Nuclear Information System (INIS)

Rajan Babu, V.; Vijayashree, R.; Govindarajan, S.; Vaidyanathan, G.; Muralikrishna, G.; Shanmugam, T.K.; Chetal, S.C.; Raghavan, K.; Bhoje, S.B.

1996-01-01

This paper presents the overall design philosophy of shutdown system of 500 MWe Prototype Fast Breeder Reactor (PFBR). It discusses design criteria, parameters calling for safety action, different safety actions and the concepts conceived for shutdown systems. In tune with the philosophy of defence-in-depth, additional passive shutdown features, viz., Self Actuating Device (SADE) and Curie Point Magnetic (CPM) switch and protective feature like absorber rod Stroke Limiting Device (SLD) are contemplated. It also discusses about suitability of Gas Expansion Module (GEM) as one of the safety devices in PFBR. (author). 3 refs, 3 figs, 1 tab

A probabilistic approach to safety/reliability of space nuclear power systems

International Nuclear Information System (INIS)

Medford, G.; Williams, K.; Kolaczkowski, A.

1989-01-01

An ongoing effort is investigating the feasibility of using probabilistic risk assessment (PRA) modeling techniques to construct a living model of a space nuclear power system. This is being done in conjunction with a traditional reliability and survivability analysis of the SP-100 space nuclear power system. The initial phase of the project consists of three major parts with the overall goal of developing a top-level system model and defining initiating events of interest for the SP-100 system. The three major tasks were performing a traditional survivability analysis, performing a simple system reliability analysis, and constructing a top-level system fault-tree model. Each of these tasks and their interim results are discussed in this paper. Initial results from the study support the conclusion that PRA modeling techniques can provide a valuable design and decision-making tool for space reactors. The ability of the model to rank and calculate relative contributions from various failure modes allows design optimization for maximum safety and reliability. Future efforts in the SP-100 program will see data development and quantification of the model to allow parametric evaluations of the SP-100 system. Current efforts have shown the need for formal data development and test programs within such a modeling framework

The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1998-01-01

This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

The EBR-II Probabilistic Risk Assessment: lessons learned regarding passive safety

Energy Technology Data Exchange (ETDEWEB)

Hill, D J; Ragland, W A; Roglans, J

1998-11-01

This paper summarizes the results from the EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10{sup -6} yr{sup -1}, even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10{sup -6} yr{sup -1} and the contribution of seismic events is 1.7 10{sup -5} yr{sup -1}. Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability.

The EBR-II probabilistic risk assessment lessons learned regarding passive safety

International Nuclear Information System (INIS)

Hill, D.J.; Ragland, W.A.; Roglans, J.

1994-01-01

This paper summarizes the results from the recently completed EBR-II Probabilistic Risk Assessment (PRA) and provides an analysis of the source of risk of the operation of EBR-II from both internal and external initiating events. The EBR-II PRA explicitly accounts for the role of reactivity feedbacks in reducing fuel damage. The results show that the expected core damage frequency from internal initiating events at EBR-II is very low, 1.6 10 -6 yr -1 , even with a wide definition of core damage (essentially that of exceeding Technical Specification limits). The annual frequency of damage, primarily due to liquid metal fires, from externally initiated events (excluding earthquakes) is 3.6 10 -6 yr -1 and the contribution of seismic events is 1.7 10 -5 yr -1 . Overall these results are considerably better than results for other research reactors and the nuclear industry in general and stem from three main sources: low likelihood of loss of coolant due to low system pressure and top entry double vessels; low likelihood of loss of decay heat removal due to reliance on passive means; and low likelihood of power/flow mismatch due to both passive feedbacks and reliability of rod scram capability

The Prenylated Rab GTPase Receptor PRA1.F4 Contributes to Protein Exit from the Golgi Apparatus.

Science.gov (United States)

Lee, Myoung Hui; Yoo, Yun-Joo; Kim, Dae Heon; Hanh, Nguyen Hong; Kwon, Yun; Hwang, Inhwan

2017-07-01

Prenylated Rab acceptor1 (PRA1) functions in the recruitment of prenylated Rab proteins to their cognate organelles. Arabidopsis ( Arabidopsis thaliana ) contains a large number of proteins belonging to the AtPRA1 family. However, their physiological roles remain largely unknown. Here, we investigated the physiological role of AtPRA1.F4, a member of the AtPRA1 family. A T-DNA insertion knockdown mutant of AtPRA1.F4 , atpra1.f4 , was smaller in stature than parent plants and possessed shorter roots, whereas transgenic plants overexpressing HA:AtPRA1.F4 showed enhanced development of secondary roots and root hairs. However, both overexpression and knockdown plants exhibited increased sensitivity to high-salt stress, lower vacuolar Na + /K + -ATPase and plasma membrane ATPase activities, lower and higher pH in the vacuole and apoplast, respectively, and highly vesiculated Golgi apparatus. HA:AtPRA1.F4 localized to the Golgi apparatus and assembled into high-molecular-weight complexes. atpra1.f4 plants displayed a defect in vacuolar trafficking, which was complemented by low but not high levels of HA : AtPRA1.F4 Overexpression of HA:AtPRA1.F4 also inhibited protein trafficking at the Golgi apparatus, albeit differentially depending on the final destination or type of protein: trafficking of vacuolar proteins, plasma membrane proteins, and trans-Golgi network (TGN)-localized SYP61 was strongly inhibited; trafficking of TGN-localized SYP51 was slightly inhibited; and trafficking of secretory proteins and TGN-localized SYP41 was negligibly or not significantly inhibited. Based on these results, we propose that Golgi-localized AtPRA1.F4 is involved in the exit of many but not all types of post-Golgi proteins from the Golgi apparatus. Additionally, an appropriate level of AtPRA1.F4 is crucial for its function at the Golgi apparatus. © 2017 American Society of Plant Biologists. All Rights Reserved.

The use of digital computers in CANDU shutdown systems

International Nuclear Information System (INIS)

Gilbert, R.S.; Komorowski, C.W.

1986-01-01

This paper summarizes the application of computers in CANDU shutdown systems. A general description of systems that are already in service is presented along with a description of a fully computerized shutdown system which is scheduled to enter service in 1987. In reviewing the use of computers in the shutdown systems there are three functional areas where computers have been or are being applied. These are (i) shutdown system monitoring, (ii) parameter display and testing and (iii) shutdown initiation. In recent years various factors (References 1 and 2) have influenced the development and deployment of systems which have addressed two of these functions. At the present time a system is also being designed which addresses all of these areas in a comprehensive manner. This fully computerized shutdown system reflects the previous design, and licensing experience which was gained in earlier applications. Prior to describing the specific systems which have been designed a short summary of CANDU shutdown system characteristics is presented

Reliability analysis of shutdown system

International Nuclear Information System (INIS)

Kumar, C. Senthil; John Arul, A.; Pal Singh, Om; Suryaprakasa Rao, K.

2005-01-01

This paper presents the results of reliability analysis of Shutdown System (SDS) of Indian Prototype Fast Breeder Reactor. Reliability analysis carried out using Fault Tree Analysis predicts a value of 3.5 x 10 -8 /de for failure of shutdown function in case of global faults and 4.4 x 10 -8 /de for local faults. Based on 20 de/y, the frequency of shutdown function failure is 0.7 x 10 -6 /ry, which meets the reliability target, set by the Indian Atomic Energy Regulatory Board. The reliability is limited by Common Cause Failure (CCF) of actuation part of SDS and to a lesser extent CCF of electronic components. The failure frequency of individual systems is -3 /ry, which also meets the safety criteria. Uncertainty analysis indicates a maximum error factor of 5 for the top event unavailability

40 CFR 180.1200 - Pseudomonas fluorescens strain PRA-25; temporary exemption from the requirement of a tolerance.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 23 2010-07-01 2010-07-01 false Pseudomonas fluorescens strain PRA-25... RESIDUES IN FOOD Exemptions From Tolerances § 180.1200 Pseudomonas fluorescens strain PRA-25; temporary... established for residues of the microbial pesticide, pseudomonas fluorescens strain PRA-25 when used on peas...

Significance of earthquake risk in nuclear power plant probabilistic risk assessments

International Nuclear Information System (INIS)

Sues, R.H.; Amico, P.J.; Campbell, R.D.

1990-01-01

During the last eight years, approximately 25 utility-sponsored probabilistic risk assessments (PRAs) have been conducted for US nuclear reactors. Of these, ten have been published, seven of which have included complete seismic risk assessment. The results of the seven published PRAs are reviewed here in order to ascertain the significance of the risk due to earthquake initiating events. While PRA methodology has been in a state of development over the past seven years, and the results are subject to interpretation (as discussed in the paper), from the review conducted it is clear that earthquake-induced initiating events are important risk contributors. It is concluded that earthquake initiating events should not be dismissed, a priori, in any nuclear plant risk assessment. (orig.)

Plasma shutdown device

International Nuclear Information System (INIS)

Hosogane, Nobuyuki; Nakayama, Takahide.

1985-01-01

Purpose: To prevent concentration of plasma currents to the plasma center upon plasma shutdown in a torus type thermonuclear device by the injection of fuels to the plasma center thereby prevent plasma disruption at the plasma center. Constitution: The plasma shutdown device comprises a plasma current measuring device that measures the current distribution of plasmas confined within a vacuum vessel and outputs a control signal for cooling the plasma center when the plasma currents concentrate to the plasma center and a fuel supply device that supplies fuels to the plasma center for cooling the center. The fuels are injected in the form of pellets into the plasmas. The direction and the velocity of the injection are set such that the pellets are ionized at the center of the plasmas. (Horiuchi, T.)

Modeling and Quantification of Team Performance in Human Reliability Analysis for Probabilistic Risk Assessment

Energy Technology Data Exchange (ETDEWEB)

Jeffrey C. JOe; Ronald L. Boring

2014-06-01

Probabilistic Risk Assessment (PRA) and Human Reliability Assessment (HRA) are important technical contributors to the United States (U.S.) Nuclear Regulatory Commission’s (NRC) risk-informed and performance based approach to regulating U.S. commercial nuclear activities. Furthermore, all currently operating commercial NPPs in the U.S. are required by federal regulation to be staffed with crews of operators. Yet, aspects of team performance are underspecified in most HRA methods that are widely used in the nuclear industry. There are a variety of "emergent" team cognition and teamwork errors (e.g., communication errors) that are 1) distinct from individual human errors, and 2) important to understand from a PRA perspective. The lack of robust models or quantification of team performance is an issue that affects the accuracy and validity of HRA methods and models, leading to significant uncertainty in estimating HEPs. This paper describes research that has the objective to model and quantify team dynamics and teamwork within NPP control room crews for risk informed applications, thereby improving the technical basis of HRA, which improves the risk-informed approach the NRC uses to regulate the U.S. commercial nuclear industry.

A methodology for Level 2 PSA evaluation with consideration of specific features for Low Power Shutdown Probabilistic Safety Assessment

Energy Technology Data Exchange (ETDEWEB)

Kim, Jae Gab; Seok, Ho [KEPCO-ENC, Yongin (Korea, Republic of)

2015-05-15

The primary objective of the Level 2 PSA during Lower Power/Shutdown (LPSD) operation is to provide insights into potential plant vulnerabilities with regard to accident progression. The shutdown risk information can be used to provide the information to develop outage risk management guidelines. The LPSD Level 2 analysis utilizes much of the at-power Level 2 analysis for bounding, conservative treatment of severe accident phenomena. But, for some portions of the analysis including Plant Operational States (POSs), LPSD-specific evaluations such as UPC related to the containment Equipment Hatch (E/H) with 4 bolts, Reactor Coolant System (RCS) Not Intact for severe accident phenomena are desired for realistic evaluation. All POSs are evaluated for their Large Release Frequency (LRF). Some POSs are evaluated conservatively utilizing the at-power models, and other POSs are evaluated in specific analysis. The overall LPSD Level 2 model is evaluated. If the containment E/H and one of the two doors on each of the personal air locks are closed as containment is operable at reduced RCS inventory operation, LRF is expected to be less than 10% of LPSD CDF.

Probabilistic safety assessment applications and insights

International Nuclear Information System (INIS)

Hitchler, M.J.; Burns, N.L.; Liparulo, N.J.; Mink, F.J.

1987-01-01

The insights gained through a comparison of seven PRA studies (Italian PUN, Sizewell B, Ringhals 2, Millstone 3, Zion 1 and 2, Oconee 3, and Seabrook) included insights regarding the adequacy of the PRA technology utilized in the studies and the potential areas for improvement and insights regarding the adequacy of plant designs and how PRA has been utilized to enhance the design and operation of nuclear power plants. (orig.)

Probabilistic risk assessment framework for structural systems under multiple hazards using Bayesian statistics

Energy Technology Data Exchange (ETDEWEB)

Kwag, Shinyoung [North Carolina State University, Raleigh, NC 27695 (United States); Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Gupta, Abhinav, E-mail: agupta1@ncsu.edu [North Carolina State University, Raleigh, NC 27695 (United States)

2017-04-15

Highlights: • This study presents the development of Bayesian framework for probabilistic risk assessment (PRA) of structural systems under multiple hazards. • The concepts of Bayesian network and Bayesian inference are combined by mapping the traditionally used fault trees into a Bayesian network. • The proposed mapping allows for consideration of dependencies as well as correlations between events. • Incorporation of Bayesian inference permits a novel way for exploration of a scenario that is likely to result in a system level “vulnerability.” - Abstract: Conventional probabilistic risk assessment (PRA) methodologies (USNRC, 1983; IAEA, 1992; EPRI, 1994; Ellingwood, 2001) conduct risk assessment for different external hazards by considering each hazard separately and independent of each other. The risk metric for a specific hazard is evaluated by a convolution of the fragility and the hazard curves. The fragility curve for basic event is obtained by using empirical, experimental, and/or numerical simulation data for a particular hazard. Treating each hazard as an independently can be inappropriate in some cases as certain hazards are statistically correlated or dependent. Examples of such correlated events include but are not limited to flooding induced fire, seismically induced internal or external flooding, or even seismically induced fire. In the current practice, system level risk and consequence sequences are typically calculated using logic trees to express the causative relationship between events. In this paper, we present the results from a study on multi-hazard risk assessment that is conducted using a Bayesian network (BN) with Bayesian inference. The framework can consider statistical dependencies among risks from multiple hazards, allows updating by considering the newly available data/information at any level, and provide a novel way to explore alternative failure scenarios that may exist due to vulnerabilities.

Probabilistic risk assessment framework for structural systems under multiple hazards using Bayesian statistics

International Nuclear Information System (INIS)

Kwag, Shinyoung; Gupta, Abhinav

2017-01-01

Highlights: • This study presents the development of Bayesian framework for probabilistic risk assessment (PRA) of structural systems under multiple hazards. • The concepts of Bayesian network and Bayesian inference are combined by mapping the traditionally used fault trees into a Bayesian network. • The proposed mapping allows for consideration of dependencies as well as correlations between events. • Incorporation of Bayesian inference permits a novel way for exploration of a scenario that is likely to result in a system level “vulnerability.” - Abstract: Conventional probabilistic risk assessment (PRA) methodologies (USNRC, 1983; IAEA, 1992; EPRI, 1994; Ellingwood, 2001) conduct risk assessment for different external hazards by considering each hazard separately and independent of each other. The risk metric for a specific hazard is evaluated by a convolution of the fragility and the hazard curves. The fragility curve for basic event is obtained by using empirical, experimental, and/or numerical simulation data for a particular hazard. Treating each hazard as an independently can be inappropriate in some cases as certain hazards are statistically correlated or dependent. Examples of such correlated events include but are not limited to flooding induced fire, seismically induced internal or external flooding, or even seismically induced fire. In the current practice, system level risk and consequence sequences are typically calculated using logic trees to express the causative relationship between events. In this paper, we present the results from a study on multi-hazard risk assessment that is conducted using a Bayesian network (BN) with Bayesian inference. The framework can consider statistical dependencies among risks from multiple hazards, allows updating by considering the newly available data/information at any level, and provide a novel way to explore alternative failure scenarios that may exist due to vulnerabilities.

Mutation of praR in Rhizobium leguminosarum enhances root biofilms, improving nodulation competitiveness by increased expression of attachment proteins.

Science.gov (United States)

Frederix, Marijke; Edwards, Anne; Swiderska, Anna; Stanger, Andrew; Karunakaran, Ramakrishnan; Williams, Alan; Abbruscato, Pamela; Sanchez-Contreras, Maria; Poole, Philip S; Downie, J Allan

2014-08-01

In Rhizobium leguminosarum bv. viciae, quorum-sensing is regulated by CinR, which induces the cinIS operon. CinI synthesizes an AHL, whereas CinS inactivates PraR, a repressor. Mutation of praR enhanced biofilms in vitro. We developed a light (lux)-dependent assay of rhizobial attachment to roots and demonstrated that mutation of praR increased biofilms on pea roots. The praR mutant out-competed wild-type for infection of pea nodules in mixed inoculations. Analysis of gene expression by microarrays and promoter fusions revealed that PraR represses its own transcription and mutation of praR increased expression of several genes including those encoding secreted proteins (the adhesins RapA2, RapB and RapC, two cadherins and the glycanase PlyB), the polysaccharide regulator RosR, and another protein similar to PraR. PraR bound to the promoters of several of these genes indicating direct repression. Mutations in rapA2, rapB, rapC, plyB, the cadherins or rosR did not affect the enhanced root attachment or nodule competitiveness of the praR mutant. However combinations of mutations in rapA, rapB and rapC abolished the enhanced attachment and nodule competitiveness. We conclude that relief of PraR-mediated repression determines a lifestyle switch allowing the expression of genes that are important for biofilm formation on roots and the subsequent initiation of infection of legume roots. © 2014 The Authors. Molecular Microbiology published by John Wiley & Sons Ltd.

Reactor shutdown system of prototype fast breeder reactor

International Nuclear Information System (INIS)

Govindarajan, S.; Singh, Om Pal; Kasinathan, N.; Paramasivan Pillai, C.; Arul, A.J.; Chetal, S.C.

2002-01-01

Full text: The shutdown system of PFBR is designed to assure a very high reliability by employing well known principles of redundancy, diversity and independence. The failure probability of the shutdown system limited to -6 / ry. Salient features of the shutdown system are: Two independent shutdown systems, each of them able to accommodate an additional single failure and made up of a trip system and an associated absorber rod group. Diversity between trip systems, rods and mechanisms. Initiation of SCRAM by two diverse physical parameters of the two shutdown systems for design events leading potentially to unacceptable conditions is the core. The first group of nine rods called control and safety rods (CSR) is used for both shutdown as well as power regulation. The second group consisting of three rods known as diverse safety rods (DSR) is used only for shutdown. Diversity between the two groups is ensured by varying the operating conditions of the electromagnets and the configurations of the mobile parts. The reactivity worth of the absorber rods have been chosen such that each group of rods would ensure cold shutdown on SCRAM even when the most reactive rod of the group fails to drop. Together the two groups ensure a shutdown margin of 5000 pcm. The speed and individual rod worth of the CSR is chosen from operational and safety considerations during reactor start up and raising of power. Required drop time of rods during SCRAM depends on the incident considered. For a severe reactivity incident of 3 $/s this has to be limited to 1s and is ensured by limiting electromagnet response time and facilitating drop by gravity. Design safety limits for core components have been determined and SCRAM parameters have been identified by plant dynamic analysis to restrict the temperatures of core components within the limits. The SCRAM parameters are distributed between the two systems appropriately. Fault tree analysis of the system has been carried out to determine the

MAAP4.0.7 analysis and justification for PRA level 1 mission success criteria

International Nuclear Information System (INIS)

Butler, J.S.; Kapitz, D.; Martin, R.P.; Seifaee, F.; Sundaram, R.K.

2008-01-01

The U.S. EPR is a 4590 MWth evolutionary pressurized water reactor that incorporates proven technology with innovative system architecture to provide an unprecedented level of safety. One of the measures of safety is provided by Probability Risk Assessment (PRA). PRA Level 1 concerns the evaluation of core damage frequency based on various initiating events and the success or failure of various plant event mitigation features. Determination of this measure requires mission success criteria, which are used to build the logic that makes up the fault trees and event trees of the Level 1 PRA. Developing mission success criteria for the wide variety of accident sequences modeled in the PRA Level 1 model requires a large number of thermal hydraulic calculations. The MAAP4 code, developed by Fauske and Associates, Inc. and distributed by EPRI, was chosen to perform these calculations because of its fast computation times relative to more sophisticated thermal-hydraulics codes This is a unique application of MAAP4, which was developed specifically for severe accident and PRA Level 2 analysis. As such, a study was performed to assess MAAP4 's thermal-hydraulic response capabilities against AREVA 's S-RELAP5 best-estimate integral systems thermal-hydraulic analysis code. (authors)

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C System

International Nuclear Information System (INIS)

Jang, Seung Cheol; Kang, Hyun Gook; Lim, Ho Gon; Park, Jin Hee; Kang, Dae Il; Eom, Heung Sub; Kim, Man Cheol; Lee, Ho Joong; Kim, Jae Whan; Ha, Jae Joo

2007-06-01

There are two technical areas to deal with in the project: the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: The LPSD PSA Area Ο Quality improvement of the KSNP LPSD PSA model in the following four technical areas; human reliability analysis (HR), system analysis (SY), data analysis (DA) and accident sequence quantification (QU) Ο Development of the LPSD configuration risk management(CRM) model - Study on the methodology for developing a CRM model, so-called ASLOC (Autonomous Shutdown LOgic Creation) - Development of the LPSD CRM model for the units of Ulchin 3 and 4 The Digital I and C PSA Area Ο Development of impact model of ESF-CCS on plant risks - Unavailability analysis of ESF-CCS for APR-1400 - Digital plant risk models for evaluating core damage frequency (CDF) Ο Study on the methodologies for treating digital-specific problems in the digital I and C PSA - Study on the methodology for evaluating safety-critical SW reliability by BBN techniques, including a feasibility study of reliability growth model - Study on the methodology for the safety-critical network system by Markov chain

Probabilistic risk assessment modeling of digital instrumentation and control systems using two dynamic methodologies

Energy Technology Data Exchange (ETDEWEB)

Aldemir, T., E-mail: aldemir.1@osu.ed [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Guarro, S. [ASCA, Inc., 1720 S. Catalina Avenue, Suite 220, Redondo Beach, CA 90277-5501 (United States); Mandelli, D. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Kirschenbaum, J. [Ohio State University, Department of Computer Science and Engineering, Columbus, OH 43210 (United States); Mangan, L.A. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Bucci, P. [Ohio State University, Department of Computer Science and Engineering, Columbus, OH 43210 (United States); Yau, M. [ASCA, Inc., 1720 S. Catalina Avenue, Suite 220, Redondo Beach, CA 90277-5501 (United States); Ekici, E. [Ohio State University, Department of Electrical and Computer Engineering, Columbus, OH 43210 (United States); Miller, D.W.; Sun, X. [Ohio State University, Nuclear Engineering Program, Columbus, OH 43210 (United States); Arndt, S.A. [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

2010-10-15

The Markov/cell-to-cell mapping technique (CCMT) and the dynamic flowgraph methodology (DFM) are two system logic modeling methodologies that have been proposed to address the dynamic characteristics of digital instrumentation and control (I and C) systems and provide risk-analytical capabilities that supplement those provided by traditional probabilistic risk assessment (PRA) techniques for nuclear power plants. Both methodologies utilize a discrete state, multi-valued logic representation of the digital I and C system. For probabilistic quantification purposes, both techniques require the estimation of the probabilities of basic system failure modes, including digital I and C software failure modes, that appear in the prime implicants identified as contributors to a given system event of interest. As in any other system modeling process, the accuracy and predictive value of the models produced by the two techniques, depend not only on the intrinsic features of the modeling paradigm, but also and to a considerable extent on information and knowledge available to the analyst, concerning the system behavior and operation rules under normal and off-nominal conditions, and the associated controlled/monitored process dynamics. The application of the two methodologies is illustrated using a digital feedwater control system (DFWCS) similar to that of an operating pressurized water reactor. This application was carried out to demonstrate how the use of either technique, or both, can facilitate the updating of an existing nuclear power plant PRA model following an upgrade of the instrumentation and control system from analog to digital. Because of scope limitations, the focus of the demonstration of the methodologies was intentionally limited to aspects of digital I and C system behavior for which probabilistic data was on hand or could be generated within the existing project bounds of time and resources. The data used in the probabilistic quantification portion of the

40 CFR 52.271 - Malfunction, startup, and shutdown regulations.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 3 2010-07-01 2010-07-01 false Malfunction, startup, and shutdown..., startup, and shutdown regulations. (a) The following regulations are disapproved because they would permit... malfunctions and/or fail to sufficiently limit startup and shutdown exemptions to those periods where it is...

Manutenção de brinquedo em praças públicas

Directory of Open Access Journals (Sweden)

Fabio Namiki

2007-12-01

Full Text Available O artigo apresenta o jacaré, um dos brinquedos executados no âmbito do Programa Centros de Bairro, que foi responsável pela implantação de cerca de 50 praças na cidade de São Paulo entre 2002 e 2004. O conjunto dos brinquedos deste programa foi apresentado e analisado no mestrado “Manutenção de praças na cidade de São Paulo. Estudo de caso: brinquedos do programa Centros de Bairro”, segundo metodologia que pode ser também aplicada para outros componentes de uma praça e mesmo para a praça em si. Espera-se que esta metodologia sirva como instrumento para o planejamento das ações de manutenção de praças e de mobiliários urbanos de modo geral. Neste texto, são apresentadas informações (da mesma forma que seriam em um manual de uso, operação e manutenção do projeto do brinquedo, obtidas junto aos responsáveis pelo programa, em entrevista com o executor dos brinquedos e através dos desenhos e documentos produzidos para a licitação e execução das peças. São também apresentadas as informações obtidas a partir das inspeções a campo e estimativas do custo de manutenção preventiva. Frente ao custo de reposição de um brinquedo novo, os valores da manutenção nos provam a importância econômica de tais ações.

Use of Human Reliability Insights to Improve Decision-Making

International Nuclear Information System (INIS)

Julius, J. A.; Moieni, P.; Grobbelaar, J.; Kohlhepp, K.

2016-01-01

This paper describes the use of insights obtained during the development and application of human reliability analysis (HRA) as part of a probabilistic risk assessment (PRA) to support decision-making, including improvements to operations, training, and safety culture. Insights have been gained from the development and application of HRA as part of a PRA for nuclear power plants in the USA, Europe and Asia over the last two decades. These models consist of Level 1 and Level 2 PRA models of internal and external events, during full power and shutdown modes of plant operation. These insights include the use of human factors information to improve the qualitative portion of the HRA. The subsequent quantification in the HRA effectively prioritises the contributors to the unreliability of operator actions, and the process facilitates the identification of the factors that are important to the success of the operator actions. Additionally, the tools and techniques also allow for the evaluation of key assumptions and sources of uncertainty. The end results have been used to effectively support decision-making for day-to-day plant operations as well as licensing issues. HRA results have been used to provide feedback and improvements to plant procedures, operator training and other areas contributing the plant safety culture. Examples of the types of insights are presented in this paper. (author)

LMFBR self-activated shutdown systems

International Nuclear Information System (INIS)

Sowa, E.S.; Barthold, W.P.; Eggen, D.T.; Huebotter, P.R.; Josephson, J.; Pizzica, P.A.; Turski, R.B.; van Erp, J.B.

1976-01-01

Self-actuated shutdown systems (SASSs), fully contained within the dimensions of a fuel subassembly and installed in the core in judiciously chosen locations, can provide an important additional safety feature for LMFBRs. If actuated by phenomena inherent to the system and its immediate environment, these systems can contribute considerably to the total reliability of the overall plant protection system, in particular as regards protection against human error. It was shown that this type of shutdown system is capable of inserting a substantial amount of negative reactivity into the core with a relatively small impact on plant performance. Furthermore, it was shown that a coolable geometry can be maintained in LMFBRs of current design for a wide spectrum of accident initiators, and for a range of response times and insertion rates which appear to be achievable within practical design limits. Experiments showed that Curie-point-operated devices have considerable promise for application in self-actuated shutdown systems, in particular as regards meeting the requirements of testability and resettability

Relationship between regulatory issues and probabilistic risk assessments

International Nuclear Information System (INIS)

Ilberg, D.; Papazoglou, I.

1985-01-01

The objective of this study was to obtain some perspective on the characteristics and the relative number of regulatory issues that are PRA related, i.e., can be effectively addressed by plant specific PRA studies. It was also aimed at developing approaches to resolution of regulatory issues as part of plant specific PRAs. Several ongoing NRC programs include a number of safety-related issues which are applicable to operating plants. A number of these issues include aspects that strongly interact with items addressed in PRA studies. The resolution of several generic issues using PRA studies has already started. A review of over 335 issues included in three NRC programs was conducted: Generic Issue Program (GI); Systematic Evaluation Program (SEP); and TMI Action Plan (TMI). The review identifid 240 items related to PRA, 120 of which were judged to have significant effect on core damage frequency. It is believed that these items can be effectively treated in a PRA study that includes internal and external events

The 10,000-year debate

International Nuclear Information System (INIS)

Wilson, J.R.

1996-01-01

Probabilistic Risk Assessment (PRA) has developed into a respected tool within the reactor community. Now, this PRA technique is being applied to a new arena, the distant future of the nuclear waste repository. Problems are already testing the credibility of PRA

The 10,000-year debate

Energy Technology Data Exchange (ETDEWEB)

Wilson, J.R.

1996-08-01

Probabilistic Risk Assessment (PRA) has developed into a respected tool within the reactor community. Now, this PRA technique is being applied to a new arena, the distant future of the nuclear waste repository. Problems are already testing the credibility of PRA.

Risks Associated with Shutdown in PWRs

International Nuclear Information System (INIS)

Grlicarev, I.

1996-01-01

The selected set of risks associated with reactor shutdown in PWRs are outlined and discussed (e. g. outage planning, residual heat removal capability, rapid boron dilution, containment integrity, fire protection). The contribution of different outage strategies to overall core damage risk during shutdown is assessed for a particular basic outage plan. The factors which increase or minimize the probability of reactor coolant boiling or core damage are analysed. (author)

Advanced Test Reactor outage risk assessment

International Nuclear Information System (INIS)

Thatcher, T.A.; Atkinson, S.A.

1997-01-01

Beginning in 1997, risk assessment was performed for each Advanced Test Reactor (ATR) outage aiding the coordination of plant configuration and work activities (maintenance, construction projects, etc.) to minimize the risk of reactor fuel damage and to improve defense-in-depth. The risk assessment activities move beyond simply meeting Technical Safety Requirements to increase the awareness of risk sensitive configurations, to focus increased attention on the higher risk activities, and to seek cost-effective design or operational changes that reduce risk. A detailed probabilistic risk assessment (PRA) had been performed to assess the risk of fuel damage during shutdown operations including heavy load handling. This resulted in several design changes to improve safety; however, evaluation of individual outages had not been performed previously and many risk insights were not being utilized in outage planning. The shutdown PRA provided the necessary framework for assessing relative and absolute risk levels and assessing defense-in-depth. Guidelines were written identifying combinations of equipment outages to avoid. Screening criteria were developed for the selection of work activities to receive review. Tabulation of inherent and work-related initiating events and their relative risk level versus plant mode has aided identification of the risk level the scheduled work involves. Preoutage reviews are conducted and post-outage risk assessment is documented to summarize the positive and negative aspects of the outage with regard to risk. The risk for the outage is compared to the risk level that would result from optimal scheduling of the work to be performed and to baseline or average past performance

Supplementary shutdown system of 220 MWe standard PHWR in India

International Nuclear Information System (INIS)

Muktibodh, U.C.

1997-01-01

The design objective of the shutdown system is to make the reactor subcritical and hold it in that state for an extended period of time. This objective must be realised under all anticipated operational occurrences and postulated abnormal conditions even during most reactive state of the core. PHWR design criteria for shutdown stipulates requirement of two independent diverse and fast acting shutdown systems, either of which acting alone should meet the above objectives. This requirement would normally call for a large number of reactivity mechanism penetrations into the calandria. From the point of view of space availability at the reactivity mechanism area on top of calandria, for the relatively small core of 220 MWe PHWRs, and ease of maintenance realisation of the total worth by either of the shutdown systems acting alone was difficult. To overcome this engineering constraint and at the same time to satisfy the design criteria, a unique approach to meet the reactivity demands for shutdown was adopted. The reactivity requirements of the shutdown consists of fast and slow reactivity changes. For the shutdown system of 220 MWe PHWRs, the approach of realizing fast reactivity changes with dual redundant, diverse, fast acting shutdown systems aided by a slow acting shutdown system to counter delayed reactivity changes was conceived. The supplementary slow acting shutdown system is called upon to act after actuation of either of the two redundant fast acting systems and is referred to as Liquid Poison Injection System (LPIS). The system adds bulk amount of neutron poison (boric acid), equivalent to 45 mk, directly into the moderator through two nozzles in calandria using pneumatic pressure. This paper describes the design of LPIS as envisaged for the standardised 220 MWe PHWRs. (author)

Examination of Conservatism in Early/Latent Fatality Estimation in Level 3 PRA

International Nuclear Information System (INIS)

Kim, Sung-yeop; Lee, Haneol; Yim, Man-Sung

2014-01-01

Due to the computational model driven-nature of the work, there exist various sources of uncertainty in level 3 PRA. They are related with source release, environmental transport and deposition, human behavior involved in dosimetry, health effect and risk assessment. For instance, a total of 376 parameters have been considered in Probabilistic Accident Consequence Uncertainty Assessment Using COSYMA and the details on the number of parameters in each analysis are listed in Table 1. In 2012, the report of NPP accident consequence simulation was distributed by the Korean Federation for Environmental Movement (KFEM). They insisted that Kori Nuclear Power Plant (NPP) accident would lead to 48,000 early fatalities and 850,000 cancer fatalities in Busan and Hanbit NPP accident would lead to 550,000 cancer fatalities in Seoul. This report exemplifies the misuse of collective dose, that is effective dose multiplied by population and time. Even though very low effective dose is considered, collective dose could give over-conservative estimate when high population and long time period is multiplied. International Commission on Radiological Protection (ICRP) forewarned about the misuse of collective dose, in their ICRP Publication 103, such as applying it to simplified calculation of fatality and risk. As part of investigation of conservatism in early and latent fatality estimation, the existing methods of early and latent fatality calculation was reviewed and the results from the use of the existing methodology were examined in this study. The method of early and latent fatality estimation in level 3 PRA was investigated and the conservatism in the result was examined in this study. For the purpose of estimating both early and latent fatality, appropriate dose distributions among the affected population are found to be important. This study showed that large conservatism may be involved in the estimated fatality if the distribution of population dose as a function of

Examination of Conservatism in Early/Latent Fatality Estimation in Level 3 PRA

Energy Technology Data Exchange (ETDEWEB)

Kim, Sung-yeop; Lee, Haneol; Yim, Man-Sung [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

2014-10-15

Due to the computational model driven-nature of the work, there exist various sources of uncertainty in level 3 PRA. They are related with source release, environmental transport and deposition, human behavior involved in dosimetry, health effect and risk assessment. For instance, a total of 376 parameters have been considered in Probabilistic Accident Consequence Uncertainty Assessment Using COSYMA and the details on the number of parameters in each analysis are listed in Table 1. In 2012, the report of NPP accident consequence simulation was distributed by the Korean Federation for Environmental Movement (KFEM). They insisted that Kori Nuclear Power Plant (NPP) accident would lead to 48,000 early fatalities and 850,000 cancer fatalities in Busan and Hanbit NPP accident would lead to 550,000 cancer fatalities in Seoul. This report exemplifies the misuse of collective dose, that is effective dose multiplied by population and time. Even though very low effective dose is considered, collective dose could give over-conservative estimate when high population and long time period is multiplied. International Commission on Radiological Protection (ICRP) forewarned about the misuse of collective dose, in their ICRP Publication 103, such as applying it to simplified calculation of fatality and risk. As part of investigation of conservatism in early and latent fatality estimation, the existing methods of early and latent fatality calculation was reviewed and the results from the use of the existing methodology were examined in this study. The method of early and latent fatality estimation in level 3 PRA was investigated and the conservatism in the result was examined in this study. For the purpose of estimating both early and latent fatality, appropriate dose distributions among the affected population are found to be important. This study showed that large conservatism may be involved in the estimated fatality if the distribution of population dose as a function of

Experience with after-shutdown decay heat removal - BWRs and PWRs

International Nuclear Information System (INIS)

Haugh, J.J.; Mollerus, F.J.; Booth, H.R.

1992-01-01

Boiling-water reactors (BWRs) and pressurized-water reactors (PWRs) make use of residual heat removal systems (RHRSs) during reactor shutdown. RHRS operational events involving an actual loss or significant degradation of an RHRS during shutdown heat removal are often prompted or aggravated by complex, changing plant conditions and by concurrent maintenance operations. Events involving loss of coolant inventory, loss of decay heat removal capability, or inadvertent pressurization while in cold shutdown have occurred. Because fewer automatic protective fetures are operative during cold shutdowns, both prevention and termination of events depend heavily on operator action. The preservation of RHRS cooling should be an important priority in all shutdown operations, particularly where there is substantial decay heat and a reduced water inventory. 13 refs., 3 figs., 4 tabs

The implications of probabilistic risk assessment for safety policy

International Nuclear Information System (INIS)

Hayns, M.R.

1987-01-01

The use of PRA results in decision making requires a level of understanding on the part of the decision maker which is higher than that obtaining previously. The most important application of PRA lies not in the final results but in the intermediate results which refer to specific systems and operations. Such intermediate results are of great value either at the design stage or later during operation. One of the most 'visible' uses of PRA results is in comparing calculated plant risks with either proposed acceptability criteria, or with other plant, or even natural events. The capability to perform PRA has been established. Only the incorporation of PRA into the licensing process is lacking. The principal conclusions on the implications of PRA for safety policy are as follows: regardless of its state of development, PRA is the only means available for calculating public risk, being able to quantify risk is important in policy related to risk acceptability and to national energy policy. PRAs will be used to establish research and development priorities. Any hazardous plant can be treated using the same methods. More sophisticated methods will be used for solving engineering problems. (author)

Pemikiran Suksesi Dalam Politik Islam Masa Pra Modern

Directory of Open Access Journals (Sweden)

Mazro'atus Sa'adah

2016-12-01

Abstrak: Pemikiran politik Islam muncul setelah Islam melalui Nabi Muhammad SAW berhasil membentuk sebuah ummat baru, dari peralihan kekuasaan kerajaan/kesukuan kepada Nabi yang kemudian kepada umat. Nabi Muhammad dinilai berhasil dalam mengatur komunitas barunya yang dikendalikan oleh ajarannya dalam seluruh lini kehidupan. Persoalan muncul kemudian setelah beliau wafat, yang akhirnya memunculkan pemikiran tentang suksesi. Artikel ini akan membahas tentang mengapa terjadi suksesi setelah Nabi Muhammad SAW wafat, bagaimana pemikiran para tokoh politik Islam masa pra modern terkait dengan suksesi, dan apa kontribusi pemikiran suksesi ini terhadap politik Islam di Indonesia. Dengan menggunakan pendekatan sejarah, ditemukan bahwa Nabi Muhammad tidak menetapkan siapa yang akan menggantikannya, dan ketika beliau wafat (632 M, para sahabat memilih seorang pemimpin (imam/khalifah. Masa pemerintahan Abu Bakar, Umar dan Usman banyak terjadi perselisihan yang awalnya terkait kepentingan agama namun berkembang menjadi kepentingan politik. Ketika Ali bin Abi Talib diangkat sebagai khalifah, konflik politik berkepanjangan berkaitan dengan pembunuhan Usman, menjadikan timbulnya perang jamal antara Aisyah dan Ali. Pada masa ini perbedaan kepentingan aqidah dipolitisir lebih jauh menjadi sebuah kepentingan politik. Dinamika politik ini kemudian melahirkan mazhab politik Islam klasik yang terbagi dalam tiga mazhab besar yaitu Sunni, Syi'ah dan Khawarij, yang darinya muncul istilah-istilah khilafah, imamah, ahlul halli wal aqdi, bay’ah, walayah dan lain-lain. Dari ketiga mazhab politik ini, kemudian muncul ide pemikiran politik Islam yang sangat kompleks dan berkepanjangan dari para tokoh politik Islam pra modern yang banyak dipengaruhi oleh filosof Yunani. Di Indonesia, pemikiran suksesi dalam politik Islam masa pra modern ini pernah diwacanakan. Namun untuk pemilihan kepala Negara belum terealisasi mengingat Indonesia bukan Negara Islam.

Probabilistic logics and probabilistic networks

CERN Document Server

Haenni, Rolf; Wheeler, Gregory; Williamson, Jon; Andrews, Jill

2014-01-01

Probabilistic Logic and Probabilistic Networks presents a groundbreaking framework within which various approaches to probabilistic logic naturally fit. Additionally, the text shows how to develop computationally feasible methods to mesh with this framework.

Use of cut-off values as meaningfulness limits in probabilistic studies and its effect on NPPs risk assessment and safety improvement

International Nuclear Information System (INIS)

Petrangeli, G.; Valeri, A.; Zaffiro, C.

1991-01-01

This paper discusses the use of cut-off values in probabilistic risk assessment/probabilistic safety assessment (PRA/PSA) of nuclear power plants (NPPs), in order to explore under which conditions this practice may help improve the meaningfulness of the results of the analyses and safety of plants, and how it may affect the assessment of risk. Reference is made, in particular, to some past practical applications, also taken from the experience of the authors within the frame of the Italian licensing process. The paper describes the Italian probabilistic criteria which use probabilistic targets and cut-off values to assess safety and identify plant safety improvements. The rationale of the approach is also discussed in the paper and results of sample applications are illustrated. The paper concludes that the use of cut-off values, if properly implemented, could be productive to improve the plant safety as it helps the analyst to focus on a restricted field of analysis, ignoring lower probability and less known events. It also points out that cut-off values should be considered as living numbers to be lowered and even eliminated as soon as significant advancements are made, through research and operational experience, in the knowledge of the pertinent events

Preliminary Evaluation of Removing Used Nuclear Fuel From Nine Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul

2013-04-30

The Blue Ribbon Commission on America’s Nuclear Future identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses. In this report, a preliminary evaluation of removing used nuclear fuel from nine shutdown sites was conducted. The shutdown sites included Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, and Zion. At these sites a total of 7649 used nuclear fuel assemblies and a total of 2813.2 metric tons heavy metal (MTHM) of used nuclear fuel are contained in 248 storage canisters. In addition, 11 canisters containing greater-than-Class C (GTCC) low-level radioactive waste are stored at these sites. The evaluation was divided in four components: • characterization of the used nuclear fuel and GTCC low-level radioactive waste inventory at the shutdown sites • an evaluation of the onsite transportation conditions at the shutdown sites • an evaluation of the near-site transportation infrastructure and experience relevant to the shipping of transportation casks containing used nuclear fuel from the shutdown sites • an evaluation of the actions necessary to prepare for and remove used nuclear fuel and GTCC low-level radioactive waste from the shutdown sites. Using these evaluations the authors developed time sequences of activities and time durations for removing the used nuclear fuel and GTCC low-level radioactive waste from a single shutdown site, from three shutdown sites located close to each other, and from all nine shutdown sites.

Safety analysis of Ignalina NPP during shutdown conditions

International Nuclear Information System (INIS)

Kaliatka, A.; Uspuras, E.

2000-01-01

The accident analysis for the Ignalina NPP with RBMK-1500 reactors at normal operating conditions and at minimum controlled power level (during startup of the reactor) has been performed in the frame of the project I n-Depth Safety Assessment of the Ignalina NPP , which was completed in 1996. However, the plant conditions during the reactor shutdown differ from conditions during reactor operation at full power (equipment status in protection systems, set points for actuation of safety and protection systems, etc.). Results of RELAP5 simulation of two worst initiating events during reactor shutdown - Pressure Header rupture in case of steam reactor cooldown as well as Pressure Header rupture in case of water reactor cooldown are discussed in the paper. Results of analysis shown that reactor are reliably cooled in both cases. Further analysis for all range of initial events during reactor shutdown and at shutdown conditions is recommended. (author)

First LHC Shutdown: Coordination and Schedule Issues

CERN Document Server

Coupard, J; Grillot, S

2010-01-01

The first LHC shutdown started in fall 2008, just after the incident on the 19th of September 2008. In addition to the typical work of a shutdown, a large number of interventions, related to the “consolidation after the incident” were performed in the LHC loop. Moreover the amount of work increased during the shutdown, following the recommendations and conclusions of the different working groups in charge of the safety of the personnel and of the machine. This paper will give an overview of the work performed, the organization of the coordination, emphasizing the new safety risks (electrical and cryogenic), and how the interventions were implemented in order to ensure both the safety of personnel and a minimized time window.

Urgensi Pemeriksaan Psikis Pra-Nikah (Studi Pandangan Kepala KUA dan Psikolog Kota Malang

Directory of Open Access Journals (Sweden)

Ika Kurnia Fitriani

2015-06-01

Full Text Available Beberapa negara muslim memberikan perhatian terhadap pemeriksaan psikis pra-nikah bagi calon mempelai, sebagai upaya menanggulangi masalah rumah tangga akibat gangguan kejiwaan di masa yang akan datang. Penelitian ini bertujuan menggali informasi dari Kepala KUA dan Psikolog di Kota Malang tentang pemeriksaan psikis pra-nikah dan urgensinya bagi calon mempelai. Penelitian ini termasuk dalam penelitian lapangan (field reasearch, dengan menggunakan pendekatan kualitatif.  Alanisis data dilakukan melalui tiga tahapan yaitu reduksi data, penyajian data, dan menarik kesimpulan. Pengecekan keabsahan data menggunakan triangulasi sumber yang membandingkan hasil wawancara dengan data sekunder, dan triangulasi teori. Hasil dari penelitian ini menunjukkan bahwa Kepala KUA dan Psikolog di kota Malang menyetujui diadakan pemeriksaan psikis pranikah akan tetapi harus ada aturan hukumnya dan dilakukan sosialisasi agar program menjadi efektif. Selain itu, pemeriksaan psikis pra-nikah tidak bertentangan dengan konsep maqashid al-syari’ah dan konsep sadz al-dzari’ah dalam hukum Islam.

Evolution of shutdown mechanism for PHWRs

International Nuclear Information System (INIS)

Singh, Manjit; Govindarajan, G.

1997-01-01

In 500 MWe PHWR, there are two independent fast acting shutdown systems namely (1) mechanical shut-off rod system and (2) liquid poison injection system. Both systems are independently capable of keeping the reactor in sub-critical condition during long shutdown. Mechanical shut-off rod system being primary shutdown system calls for a very high reliability of operation as well as effectiveness, which are mainly governed by its ability to operate within a very short time and the magnitude of negative reactivity worth it can provide. Mechanical shut-off rods are normally parked above the core by shut-off rod drive mechanism. On receiving a scram signal, shut-off rods are released from the holding electromagnetic clutch and fall under gravity into the core. This paper discusses the salient features of mechanical shut-off rod system. A brief account of detailed design and development of sub-assemblies of shut-off rod drive mechanism is also presented. (author)

The Use and Development of Probabilistic Safety Assessment in NEA Member Countries

International Nuclear Information System (INIS)

2002-01-01

The mission of the CSNI is to assist Member countries in maintaining and further developing the scientific and technical knowledge base required to assess the safety of nuclear reactors and fuel cycle facilities. The mission of the Working Group on Risk Assessment (WGRisk) is to advance the understanding and utilisation of Probabilistic Safety Assessment (PSA) in ensuring continued safety of nuclear installations in Member countries. In pursuing this goal, the Working Group shall recognize the different methodologies for identifying contributors to risk and assessing their importance. While the Working Group shall continue to focus on the more mature PSA methodologies for Level 1, Level 2, internal, external, shutdown, etc. It shall also consider the applicability and maturity of PSA methods for considering evolving issues such as human reliability, software reliability, ageing issues, etc., as appropriate. This report provides descriptions of the current status of PSA programmes in Member countries including basic background information, guidelines, various PSA applications, major results in recent studies, PSA based plant modifications and research and development topics. While the compilation is a not complete compilation it provides a 'snapshot' of the current situation in the Member countries and hence it provides reference information and various insights to both the PSA practician and others involved in the nuclear industry. The terms PSA (Probabilistic Safety Assessment) and PRA (Probabilistic Risk Assessment) are utilised to denote this subject. In each of the chapters the objective is to present a 'snapshot' of the current status. The main issues considered in the different chapters are Background Information, Quantitative Safety Guidelines, Status of PSA Programmes, PSA Applications, PSA Related Research and Development and PSA Based Plant Modifications. It is important to note that the information contained in this report represents current practices in

2009 Space Shuttle Probabilistic Risk Assessment Overview

Science.gov (United States)

Hamlin, Teri L.; Canga, Michael A.; Boyer, Roger L.; Thigpen, Eric B.

2010-01-01

Loss of a Space Shuttle during flight has severe consequences, including loss of a significant national asset; loss of national confidence and pride; and, most importantly, loss of human life. The Shuttle Probabilistic Risk Assessment (SPRA) is used to identify risk contributors and their significance; thus, assisting management in determining how to reduce risk. In 2006, an overview of the SPRA Iteration 2.1 was presented at PSAM 8 [1]. Like all successful PRAs, the SPRA is a living PRA and has undergone revisions since PSAM 8. The latest revision to the SPRA is Iteration 3. 1, and it will not be the last as the Shuttle program progresses and more is learned. This paper discusses the SPRA scope, overall methodology, and results, as well as provides risk insights. The scope, assumptions, uncertainties, and limitations of this assessment provide risk-informed perspective to aid management s decision-making process. In addition, this paper compares the Iteration 3.1 analysis and results to the Iteration 2.1 analysis and results presented at PSAM 8.

Estrogen and progesterone receptors have distinct roles in the establishment of the hyperplastic phenotype in PR-A transgenic mice

Energy Technology Data Exchange (ETDEWEB)

Simian, Marina; Bissell, Mina J.; Barcellos-Hoff, Mary Helen; Shyamala, Gopalan

2009-05-11

Expression of the A and B forms of progesterone receptor (PR) in an appropriate ratio is critical for mammary development. Mammary glands of PR-A transgenic mice, carrying an additional A form of PR as a transgene, exhibit morphological features associated with the development of mammary tumors. Our objective was to determine the roles of estrogen (E) and progesterone (P) in the genesis of mammary hyperplasias/preneoplasias in PR-A transgenics. We subjected PR-A mice to hormonal treatments and analyzed mammary glands for the presence of hyperplasias and used BrdU incorporation to measure proliferation. Quantitative image analysis was carried out to compare levels of latency-associated peptide and transforming growth factor beta 1 (TGF{beta}1) between PR-A and PR-B transgenics. Basement membrane disruption was examined by immunofluorescence and proteolytic activity by zymography. The hyperplastic phenotype of PR-A transgenics is inhibited by ovariectomy, and is reversed by treatment with E + P. Studies using the antiestrogen ICI 182,780 or antiprogestins RU486 or ZK 98,299 show that the increase in proliferation requires signaling through E/estrogen receptor alpha but is not sufficient to give rise to hyperplasias, whereas signaling through P/PR has little impact on proliferation but is essential for the manifestation of hyperplasias. Increased proliferation is correlated with decreased TGF{beta}1 activation in the PR-A transgenics. Analysis of basement membrane integrity showed loss of laminin-5, collagen III and collagen IV in mammary glands of PR-A mice, which is restored by ovariectomy. Examination of matrix metalloproteases (MMPs) showed that total levels of MMP-2 correlate with the steady-state levels of PR, and that areas of laminin-5 loss coincide with those of activation of MMP-2 in PR-A transgenics. Activation of MMP-2 is dependent on treatment with E and P in ovariectomized wild-type mice, but is achieved only by treatment with P in PR-A mice. These data

Innovative real time simulation training and nuclear probabilistic risk assessment

International Nuclear Information System (INIS)

Reisinger, M.F.

1991-01-01

Operator errors have been an area of public concern for the safe operation of nuclear power plants since the TMI2 incident. Simply stated, nuclear plants are very complex systems and the public is skeptical of the operators' ability to comprehend and deal with the vast indications and complexities of potential nuclear power plant events. Prior to the TMI2 incident, operator errors and human factors were not included as contributing factors in the Probabilistic Risk Assessment (PRA) studies of nuclear power plant accidents. More recent efforts in nuclear risk assessment have addressed some of the human factors affecting safe nuclear plant operations. One study found four major factors having significant impact on operator effectiveness. This paper discusses human factor PRAs, new applications in simulation training and the specific potential benefits from simulation in promoting safer operation of future power plants as well as current operating power plants

The application of probabilistic risk assessment to a LLW incinerator

International Nuclear Information System (INIS)

Li, K.K.; Huang, F.T.

1993-01-01

The 100 Kg/hr low-level radioactive waste (LLW) incinerator and the 1,500 ton supercompactor are two main vehicles in the Taiwan Power Company's Volume Reduction Center. Since the hot test of the incinerator in mid 1990, various problems associated with the original design and operating procedures were encountered. During the early stages of putting an incinerator in service, the modification and fine-tuning of the system would help future reliable operations. The probabilistic risk assessment (PRA) method was introduced to evaluate the interaction between potential system failure and its environmental impact and further help diagnose the system defects initially. The draft Level 1 system analysis was completed and the event and fault trees were constructed. Qualitatively, this approach is useful for preventing the system failure from occurring. However, Levels 2 and 3 analysis can only be done when sufficient data become available in the future

Probabilistic risk assessment and intelligent decision support systems

International Nuclear Information System (INIS)

Wu, J.S.; Apostolakis, G.E.; Okrent, D.

1989-01-01

The purpose of this paper is to review the progress made in recent years in both the area of PRA as a support to AI applications and the area of AI applications in PRA. The emphasis is on the areas that have made some progress in the past few years, with a brief description of the methods and a discussion of the potential uses and weaknesses. Also included is a brief review of recent developments in the theory of uncertainty in the AI community that may impact uncertainty modeling in PRA. AI techniques could be applied to the related field of PRA in several ways. In this discussion, however, the scope is limited to emergency diagnosis and accident management, because these are the areas that have attracted most of the attention in recent years. The potential use of PRA as a support to these applications is discussed in detail, and this is followed by a survey of recent developments in these areas. (orig./GL)

CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

International Nuclear Information System (INIS)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author)

CANDU 6 liquid injection shutdown system waterhammer analysis using PTRAN

Energy Technology Data Exchange (ETDEWEB)

Ko, Deuk Yoon; Kim, Eun Ki; Ko, Yong Sang; Park, Byung Ho; Kim, Seok Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

1996-06-01

An in-core LOCA could result in flooding of the helium header in the liquid injection shutdown system. Flooding of the helium header will result in severe pressure transients (waterhammer) in the liquid injection shutdown system when the shutdown signal is initiated. To evaluate the impact of the dynamic effects of this event, a pressure transient analysis has been performed. This analysis is performed using PTRAN, which is a computer program based on the method of characteristics. The results of this analysis are used in the stress analysis of the piping and pipe supports to ensure that the liquid injection shutdown system can withstand the pressure transient loadings. This analysis report documents the results of waterhammer analysis performed for the liquid injection shutdown system for the Wolsung nuclear power plant unit 2, 3 and 4. 4 tabs., 11 figs., 15 refs. (Author).

Applying the results of probablistic safety analysis of nuclear power plants: a survey of experience

International Nuclear Information System (INIS)

Andrews, W.B.; Herttrich, M.; Koeberlein, K.; Schwager.

1985-01-01

To date, discussions of the many different types of potential applications of PRA/PSA results and insights to safety-decision-making have been mainly theoretical. Various safety goals have been proposed as decision criteria. However, the discussion on the role of PRA/PSA and Safety Goals in safety-decision-making, especially in licensing, is controversial. A Working Group of the OECD Nuclear Energy Agency is completing a compilation and evaluation of real examples of past and present practical experience with the application of probabilistic methods in reactor safety decision-making, with the idea of developing a common understanding in this area. More than fifty different cases where PRA has influenced decision-making have been surveyed. These include, for example, regulatory changes, fixing of licensing requirements, plant specific modifications of design of operation, prioritization of safety issues and emergency planning. This feedback of experience - both positive and negative - with PRA/PSA applications is considered to provide guidance on how probabilistic approaches can be introduced into current safety practices, and on desirable future developments in probabilistic methods and specific PSA/PRA studies. Generic insights from the survey are given

Study on secondary shutdown systems in Tehran research reactor

Energy Technology Data Exchange (ETDEWEB)

Jalali, H.R.; Fadaei, A.H., E-mail: Fadaei_amir@aut.ac.ir; Gharib, M.

2015-09-15

Highlights: • A study was undertaken to summarize the techniques for secondary shutdown systems (SSS). • Neutronic calculation performed for proposed systems as SSS. • Dumping the heavy water stored in the reflector vessel is capable to shut down reactor. • Neutronic and transient calculation was done for validating the selected SSS. • All calculation shown that this system has advantages in safety and neutron economy. - Abstract: One important safety aspect of any research reactor is the ability to shut down the reactor. Usually, research reactors, currently in operation, have a single shutdown system based on the simultaneous insertion of the all control rods into the reactor core through gravity. Nevertheless, the International Atomic Energy Agency currently recommends use of two shutdown systems which are fully independent from each other to guarantee secure shutdown when one of them fails. This work presents an investigative study into secondary shutdown systems, which will be an important safety component in the research reactor and will provide another alternative way to shut down the reactor emergently. As part of this project, a study was undertaken to summarize the techniques that are currently used at world-wide research reactors for recognizing available techniques to consider in research reactors. Removal of the reflector, removal of the fuels, change in critical shape of reactor core and insertion of neutron absorber between the core and reflector are selected as possible techniques in mentioned function. In the next step, a comparison is performed for these methods from neutronic aspects. Then, chosen method is studied from the transient behavior point of view. Tehran research reactor which is a 5 MW open-pool reactor selected as a case study and all calculations are carried out for it. It has 5 control rods which serve the purpose of both reactivity control and shutdown of reactor under abnormal condition. Results indicated that heavy

Optimization method to branch-and-bound large SBO state spaces under dynamic probabilistic risk assessment via use of LENDIT scales and S2R2 sets

International Nuclear Information System (INIS)

Nielsen, Joseph; Tokuhiro, Akira; Khatry, Jivan; Hiromoto, Robert

2014-01-01

Traditional probabilistic risk assessment (PRA) methods have been developed to evaluate risk associated with complex systems; however, PRA methods lack the capability to evaluate complex dynamic systems. In these systems, time and energy scales associated with transient events may vary as a function of transition times and energies to arrive at a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. Unfortunately DPRA methods introduce issues associated with combinatorial explosion of states. In order to address this combinatorial complexity, a branch-and-bound optimization technique is applied to the DPRA formalism to control the combinatorial state explosion. In addition, a new characteristic scaling metric (LENDIT – length, energy, number, distribution, information and time) is proposed as linear constraints that are used to guide the branch-and-bound algorithm to limit the number of possible states to be analyzed. The LENDIT characterization is divided into four groups or sets – 'state, system, resource and response' (S2R2) – describing reactor operations (normal and off-normal). In this paper we introduce the branch-and-bound DPRA approach and the application of LENDIT scales and S2R2 sets to a station blackout (SBO) transient. (author)

Criteria for remote shutdown for light water reactors

International Nuclear Information System (INIS)

Anon.

1983-01-01

This Standard provides design criteria which require that: (1) specific controls and monitoring equipment shall be provided for achieving and maintaining the plant in a safe shutdown condition; (2) these controls be installed at a location (or locations) that is physically remote from the control room and cable spreading areas; (3) simultaneous control from both locations shall be prevented by administrative controls or devices for transfer of control from the control room to the remote location(s); and (4) the remote controls be used as defense-in-depth measure in addition to the control room shutdown controls and as a minimum shall provide for one complete channel of shutdown equipment

CAREM-25 Reactor Second Shutdown System Consolidation Analysis

International Nuclear Information System (INIS)

Gimenez, Marcelo; Zanocco, Pablo; Schlamp, Miguel

2000-01-01

CAREM Reactor Second Shutdown System (SSS) injects boron into the primary circuit in case of First Shutdown System failure in order to stop the nuclear reaction and to maintain the core in a safe condition during cold shutdown.It also has another safety function which is to inject water in the primary system at any pressure in case of LOCA.Different system requirements are analyzed during a SSS spurious trip and LOCA's transients.Two different alternatives are presented for the stand by condition pressurized system, they are solid mode and hot water layer. Both cases fulfill the design requirements from the safety point of view

Development of a methodology for post closure radiological risk analysis of underground waste repositories. Illustrative assessment of the Harwell site. V.1

International Nuclear Information System (INIS)

Gralewski, Z.A.; Kane, P.; Nicholls, D.B.

1987-06-01

A probabilistic risk analysis (pra) is demonstrated for a number of ground water mediated release scenarios at the Harwell Site for a hypothetical repository at a depth of about 150 metres. This is the second stage of development of an overall risk assessment methodology. A procedure for carrying out multi-scenario assessment using available probabilistic risk assessment (pra) models is presented and a general methodology for combining risk contributions is outlined. Appropriate levels of model complexity in pra are discussed. Modelling requirements for the treatment of multiple simultaneous pathways and of site evolution are outlined. Further developments of pra systems are required to increase the realism of both the models and their mode of application, and hence to improve estimates of risk. (author)

Development of a methodology for post closure radiological risk analysis of underground waste repositories. Illustrative assessment of the Harwell site

International Nuclear Information System (INIS)

Gralewski, Z.A.; Kane, P.; Nicholls, D.B.

1987-06-01

A probabilistic risk analysis (pra) is demonstrated for a number of ground water mediated release scenarios at the Harwell Site for a hypothetical repository at a depth of about 150 metres. This is the second stage of development of an overall risk assessment methodology. A procedure for carrying out multi-scenario assessment using available probabilistic risk assessment (pra) models is presented and a general methodology for combining risk contributions is outlined. Appropriate levels of model complexity in pra are discussed. Modelling requirements for the treatment of multiple simultaneous pathways and of site evolution are outlined. Further developments of pra systems are required to increase the realism of both the models and their mode of application, and hence to improve estimates of risk. (author)

Certificate for Safe Emergency Shutdown of Wind Turbines

DEFF Research Database (Denmark)

Wisniewski, Rafal; Svenstrup, Mikael; Pedersen, Andreas Søndergaard

2013-01-01

To avoid damage to a wind turbine in the case of a fault or a large wind gust, a detection scheme for emergency shutdown is developed. Specifically, the concept of a safety envelope is introduced. Within the safety envelope, the system can be shutdown without risking structural damage to the turb...

Probabilistic risk assessment for salt repository conceptual design of subsurface facilities: A techical basis for Q-list determination

International Nuclear Information System (INIS)

Chen, C.P.; Mayberry, J.J.; Shepherd, J.; Koza, H.; Rahmani, H.; Sinsky, J.

1987-12-01

Subpart G ''Quality Assurance'' of 10 CFR Part 60 requires that the US Department of Energy (DOE) apply a quality assurance program to ''all systems, structures, and components important to safety'' and to ''design and characterization of barriers important to waste isolation.'' In April 1986, DOE's Office of Geologic Repositories (OGR) issued general guidance for formulating a list of such systems, structures, and components---the Q-list. This guidance called for the use of probabilistic risk assessment (PRA) techniques to identify Q-list items. In this report, PRA techniques are applied to the underground facilities and systems described in the conceptual design report for the Salt Repository Project (SRP) in Deaf Smith County, Texas. Based on probability and dose consequence calculations, no specific items were identified for the Q-list. However, evaluation of the analyses indicated that two functions are important in precluding off-site releases of radioactivity: disposal container integrity; and isolation of the underground facility by the heating, ventilation, and air conditioning (HVAC) systems. Items related to these functions are recommended for further evaluation as the repository design progresses. 13 refs., 20 figs

SAPHIRE 8 Volume 3 - Users' Guide

Energy Technology Data Exchange (ETDEWEB)

C. L. Smith; K. Vedros; K. J. Kvarfordt

2011-03-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). This reference guide will introduce the SAPHIRE Version 8.0 software. A brief discussion of the purpose and history of the software is included along with general information such as installation instructions, starting and stopping the program, and some pointers on how to get around inside the program. Next, database concepts and structure are discussed. Following that discussion are nine sections, one for each of the menu options on the SAPHIRE main menu, wherein the purpose and general capabilities for each option are

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

International Nuclear Information System (INIS)

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

1992-11-01

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

Probabilistic safety assessment of WWER440 reactors prediction, quantification and management of the risk

CERN Document Server

Kovacs, Zoltan

2014-01-01

The aim of this book is to summarize probabilistic safety assessment (PSA) of nuclear power plants with WWER440 reactors and  demonstrate that the plants are safe enough for producing energy even in light of the Fukushima accident. The book examines level 1 and 2 full power, low power and shutdown PSA, and summarizes the author's experience gained during the last 35 years in this area. It provides useful examples taken from PSA training courses the author has lectured and organized by the International Atomic Energy Agency. Such training courses were organised in Argonne National Laboratory (

Evaluation of hsp65 Nested PCR-Restriction Analysis (PRA) for Diagnosing Tuberculosis in a High Burden Country

Science.gov (United States)

Macente, Sara; Fujimura Leite, Clarice Queico; Santos, Adolfo Carlos Barreto; Siqueira, Vera Lúcia Dias; Machado, Luzia Neri Cosmo; Marcondes, Nadir Rodrigues; Hirata, Mario Hiroyuki; Hirata, Rosário Dominguez Crespo

2013-01-01

Current study evaluated the hsp65 Nested PCR Restriction Fragment Length Polymorphism Analysis (hsp65 Nested PCR-PRA) to detect and identify Mycobacterium tuberculosis complex directly in clinical samples for a rapid and specific diagnosis of tuberculosis (TB). hsp65 Nested PCR-PRA was applied directly to 218 clinical samples obtained from 127 patients suspected of TB or another mycobacterial infection from July 2009 to July 2010. The hsp65 Nested PCR-PRA showed 100% sensitivity and 95.0 and 93.1% specificity in comparison with culture and microscopy (acid fast bacillus smear), respectively. hsp65 Nested PCR-PRA was shown to be a fast and reliable assay for diagnosing TB, which may contribute towards a fast diagnosis that could help the selection of appropriate chemotherapeutic and early epidemiological management of the cases which are of paramount importance in a high TB burden country. PMID:24260739

Clinical analysis of the changes of plasma PRA, AT-II and Aid levels in patients with acute renal failure

International Nuclear Information System (INIS)

Zhang Qiuyue; Yang Yongqing

2002-01-01

Objective: To investigate the role of changes of plasma PRA, AT-II and Ald levels in the pathogenesis of acute renal failure. Methods: Plasma PRA, AT-II and Ald levels were determined with RIA in 40 normal subjects and 72 cases of acute renal failure. Results: Plasma PRA, AT-II and Ald levels in the patients were markedly increased as compared with those in normal subjects (p < 0.05, p < 0.01, p < 0.001 respectively). There were no linearity and exponential relationship between plasma PRA, AT-II, Ald levels and the 24 h urinary sodium excretion amount (within the range of 89.1 - 365.2 mEq). Conclusion: Acute renal failure could activate the RAAS function

Quality of the current low power and shutdown PSA practice

International Nuclear Information System (INIS)

Jang, Seung Cheol; Park, Jin Hee; Lim, Ho Gon; Kim, Tae Woon

2004-01-01

A probabilistic safety assessment (PSA) for the low-power and shutdown (LPSD) modes in a Korea standard nuclear power plant (KSNP) has been performed for the purpose of estimating the LPSD risk and identifying the vulnerabilities of LPSD operations. Both the operational experience and PSA results indicate that the risks from LPSD operations could be comparable with those from power operations. However, the application of the LPSD risk insights to risk-informed decision making has been slow to be adopted in practice. It is largely due to the question of whether the current LPSD PSA practice is appropriate for application to risk-informed decision making or not. Such a question has to do with the quality of the current LPSD PSA practice. In this paper, we have performed self-assessment of the KSNP LPSD PSA quality based on the ANS Standard (draft as of 13 Sep. 2002). The aims of the work are to find the LPSD PSA technical areas insufficient for application to risk-informed decision making and to efficiently allocate the limited research resources to improve the LPSD PSA model quality. Many useful findings regarding the current LPSD PSA quality are presented in this paper

Safety aspects of unplanned shutdowns and trips

International Nuclear Information System (INIS)

1986-05-01

The issue of unplanned shutdowns and trips is receiving increased attention worldwide in view of its importance to plant safety and availability. There exists significant variation in the number of forced shutdowns for nuclear power plants of the same type operating worldwide. The reduction of the frequency of these events will have safety benefits in terms of reducing the frequency of plant transients and the challenges to the safety systems, and the risks of possible incidents. This report provides an insight into the causes of unplanned shutdowns experienced in operating nuclear power plants worldwide, the good practices that have been found effective in minimizing their occurrence, and the measures that have been taken to reduce these events. Specific information on the experiences, approaches and practices of some countries in dealing with this issue is presented in Appendix A

Perspectives on Low Power and Shutdown Risk

International Nuclear Information System (INIS)

Camp, Allen L.; Whitehead, Donnie W.; Wheeler, Timothy A.; Lehner, John; Chu, Tsong-Lun; Lois, Erasmai; Drouin, Mary

2000-01-01

This paper presents results from a program sponsored by the US Nuclear Regulatory Commission to examine the risks from low power and shutdown operations. Significant progress has been made by the industry in reducing such risks; however, important operational events continue to occur. Current perceptions of low power and shutdown risks are discussed in the paper along with an assessment of the current methods for understanding important events and quantifying their associated risk

An integrated PRA module for fast determination of risk significance and improvement effectiveness

International Nuclear Information System (INIS)

Chao, Chun-Chang; Lin, Jyh-Der

2004-01-01

With the widely use of PRA technology in risk-informed applications, to predict the changes of CDF and LERF becomes a standard process for risk-informed applications. This paper describes an integrated PRA module prepared for risk-informed applications. The module contains a super risk engine, a super fault tree engine, an advanced PRA model and a tool for data base maintenance. The individual element of the module also works well for purpose other than risk-informed applications. The module has been verified and validated through a series of scrupulous benchmark tests with similar software. The results of the benchmark tests showed that the module has remarkable accuracy and speed even for an extremely large-size top-logic fault tree as well as for the case in which large amount of MCSs may be generated. The risk monitor for nuclear power plants in Taiwan is the first application to adopt the module. The results predicted by the risk monitor are now accepted by the regulatory agency. A tool to determine the risk significance according to the inspection findings will be the next application to adopt the module in the near future. This tool classified the risk significance into four different color codes according to the level of increase on CDF. Experience of application showed that the flexibility, the accuracy and speed of the module make it useful in any risk-informed applications when risk indexes must be determined by resolving a PRA model. (author)

Review of the Shoreham Nuclear Power Station Probabilistic Risk Assessment: internal events and core damage frequency

International Nuclear Information System (INIS)

Ilberg, D.; Shiu, K.; Hanan, N.; Anavim, E.

1985-11-01

A review of the Probabilistic Risk Assessment of the Shoreham Nuclear Power Station was conducted with the broad objective of evaluating its risks in relation to those identified in the Reactor Safety Study (WASH-1400). The scope of the review was limited to the ''front end'' part, i.e., to the evaluation of the frequencies of states in which core damage may occur. Furthermore, the review considered only internally generated accidents, consistent with the scope of the PRA. The review included an assessment of the assumptions and methods used in the Shoreham study. It also encompassed a reevaluation of the main results within the scope and general methodological framework of the Shoreham PRA, including both qualitative and quantitative analyses of accident initiators, data bases, and accident sequences which result in initiation of core damage. Specific comparisons are given between the Shoreham study, the results of the present review, and the WASH-1400 BWR, for the core damage frequency. The effect of modeling uncertainties was considered by a limited sensitivity study so as to show how the results would change if other assumptions were made. This review provides an independently assessed point value estimate of core damage frequency and describes the major contributors, by frontline systems and by accident sequences. 17 figs., 81 tabs

Italy: Analysis of Solutions for Passively Actuated Safety Shutdown Devices

International Nuclear Information System (INIS)

Burgazzi, L.

2015-01-01

This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on Fast Reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in Sodium Fast Reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to Fast Reactor design to meet the safety requirements

Case studies: Risk-based analysis of technical specifications

International Nuclear Information System (INIS)

Wagner, D.P.; Minton, L.A.; Gaertner, J.P.

1987-01-01

The SOCRATES computer program uses the results of a Probabilistic Risk Assessment (PRA) or a system level risk analysis to calculate changes in risk due to changes in the surveillance test interval and/or the allowed outage time stated in the technical specification. The computer program can accommodate various testing strategies (such as staggered or simultaneous testing) to allow modeling of component testing as it is carried out at a plant. The methods and computer program are an integral part of a larger decision process aimed at determining benefits from technical specification changes. These benefits can include cost savings to the utilities by reducing forced shutdowns with no adverse impacts on risk. Three summaries of case study applications are included to demonstrate the types of results that can be achieved through risk-based evaluation of technical specifications. (orig.)

Reactor Shutdown Mechanism by Top-mounted Hydraulic System

Energy Technology Data Exchange (ETDEWEB)

Kim, Sang Haun; Cho, Yeong Garp; Choi, Myoung Hwan; Lee, Jin Haeng; Huh, Hyung; Kim, Jong In [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2012-05-15

There are two types of reactor shutdown mechanisms in HANARO. One is the mechanism driven by a hydraulic system, and the other is driven by a stepping motor. In HANARO, there are four Control Rod Drive Mechanisms (CRDMs) with an individual step motor and four Shutoff (SO) Units with an individual hydraulic system located at the top of reactor pool. The absorber rods in SO units are poised at the top of the core by the hydraulic force during normal operation. The rods of SO units drop by gravity as the first reactor showdown mechanism when a trip is commended by the reactor protection system (RPS). The rods in CRDMs also drop by gravity together as a redundant shutdown mechanism. When a trip is commended by the reactor regulating system (RRS), the absorber rods of CRDM only drop; while the absorber rods of SO units stay at the top of the core by the hydraulic system. The reactivity control mechanisms of in JRTR, one of the new research reactor with plate type fuels, consist of four CRDMs driven by an individual step motor and two second shutdown drive mechanisms (SSDMs) driven by an individual hydraulic system as shown in Fig. 1. The CRDMs act as the first reactor shutdown mechanism and reactor regulating as well. The top-mounted SSDM driven by the hydraulic system for the JRTR is under design in KAERI. The SSDM provides an alternate and independent means of reactor shutdown. The second shutdown rods (SSRs) of the SSDM are poised at the top of the core by the hydraulic system during the normal operation and drop by gravity for the reactor trip. Based on the proven technology of the design, operation and maintenance for HANARO, the SSDM for the JRTR has been optimized by the design improvement from the experience and test. This paper aims for the introduction of the SSDM in the process of the basic design. The major differences of the shutdown mechanisms by the hydraulic system are compared between HANARO and JRTR, and the design features, system, structure and

Survey of seismic fragilities used in PRA studies of nuclear power plants

International Nuclear Information System (INIS)

Park, Y.J.; Hofmayer, C.H.; Chokshi, N.C.

1998-01-01

In recent years, seismic PRA studies have been performed on a large number of nuclear power plants in the USA. This paper presents a summary of a survey on fragility databases and the range of evaluated fragility values of various equipment categories based on past PRAs. The survey includes the use of experience data, the interpretations of available test data, and the quantification of uncertainties. The surveyed fragility databases are limited to data available in the public domain such as NUREG reports, conference proceedings and other publicly available reports. The extent of the availability of data as well as limitations are studied and tabulated for various equipment categories. The survey of the fragility values in past PRA studies includes not only the best estimate values, but also the dominant failure modes and the estimated uncertainty levels for each equipment category. The engineering judgments employed in estimating the uncertainty in the fragility values are also studied. This paper provides a perspective on the seismic fragility evaluation procedures for equipment in order to clearly identify the engineering analysis and judgment used in past seismic PRA studies

Applicability of PRA methods and data to the financial risk assessment of nuclear power plants

International Nuclear Information System (INIS)

El-Sheik, K.A.

1985-01-01

Financial risk assessment, where the probability and severity of financial consequences are estimated, offers a logical framework for organizing and evaluating data pertinent to nuclear power plant accidents. Under the sponsorship of the Electric Power Research Institute, General Electric investigated the feasibility of financial risk assessment of nuclear power plants and of applying PRA methods and data in such an assessment. This paper summarizes the main findings of this investigation. Specifically, the paper discussed the following topics: definition of financial consequences and financial risk; overall approach for financial risk assessment and how it compares with the approach for PRA used in the Reactor Safety Study; and specific financial risk assessment procedures for defining initiating events, plant response sequences, institutional scenarios, and financial consequences and how they compare to analogous procedures for PRA

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

International Nuclear Information System (INIS)

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

1994-08-01

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

Optimization of reactor coolant shutdown chemistry practices for crud inventory management

International Nuclear Information System (INIS)

Fellers, B.; Barnette, J.; Stevens, J.; Perkins, D.

2002-01-01

This report describes reactor coolant shutdown chemistry control practices at Comanche Peak Steam Electric Station (CPSES, TXU-Generation, USA). The shutdown evolution is managed from a process control perspective to achieve conditions most favorable to crud decomposition and to avoiding re-precipitation of metals. The report discusses the evolution of current industry practices and the necessity for greater emphasis on shutdown chemistry control in response to Axial Offset Anomaly and growth of ex-core radiation fields during outage conditions. Nuclear Industry experience with axial offset anomaly (AOA), radiation field growth and unexpected behavior of crud during reactor shutdowns has encouraged the refinement of chemistry control practices during plant shutdown and startup. The strong implication of nickel rich crud as a cause of AOA and unexpected crud behavior has resulted in a focus on nickel inventory management. The goals for Comanche Peak Steam Electric Station (CPSES) include maintaining solubility of metals and radioisotopes, maximizing nickel removal and effective cleanup with demineralizers. This paper provides results and lessons learned from long term efforts to optimize the shutdown process. (authors)

Startup, Shutdown, & Malfunction (SSM) Emissions

Science.gov (United States)

EPA issued a final action to ensure states have plans in place that are fully consistent with the Clean Air Act and recent court decisions concerning startup, shutdown and malfunction (SSM) operations.

Procedures for analysis of accidents in shutdown modes for WWER nuclear power plants. A publication of the extrabudgetary programme on the safety of WWER and RBMK nuclear power plants

International Nuclear Information System (INIS)

1997-07-01

Operational events occurring during shutdown conditions contribute significantly to the NPP risk due to the fact that both preventive and mitigatory capabilities of the plant are somehow degraded. The need for detailed information in the performance and review of accident analysis for WWER type NPPs was identified as a priority within IAEA Extrabudgetary Program on Safety of WWER and RBMK NPPs. The present guidelines were developed through two consultants meetings in 1995 and 1996. The guidelines establish a set of criteria for performing deterministic analysis of accidents, initiated by events occurring under shutdown conditions. This report is mostly relevant for licensing type calculations, and may to a certain extent, also used for development, improvement or justification of the plant limits and conditions, emergency operating procedures, operator training programs and probabilistic safety studies. The guidelines apply to all WWER plants in operation and/or under construction

Stabilization and shutdown of Oak Ridge National Laboratory's Radioisotopes Production Facility

International Nuclear Information System (INIS)

Eversole, R.E.

1992-01-01

The Oak Ridge National Laboratory (ORNL) has been involved in the production and distribution of a variety of radioisotopes for medical, scientific and industrial applications since the late 1940s. Production of these materials was concentrated in a number of facilities primarily built in the 1950s and 1960s. Due to the age and deteriorating condition of these facilities, it was determined in 1989 that it would not be cost effective to upgrade these facilities to bring them into compliance with contemporary environmental, safety and health standards. The US Department of Energy (DOE) instructed ORNL to halt the production of isotopes in these facilities and maintain the facilities in safe standby condition while preparing a stabilization and shutdown plan. The goal was to place the former isotope production facilities in a radiologically and industrially safe condition to allow a 5-year deferral of the initiation of environmental restoration (ER) activities. In response to DOE's instructions, ORNL identified 17 facilities for shutdown, addressed the shutdown requirements for each facility, and prepared and implemented a three-phase, 4-year plan for shutdown of the facilities. The Isotopes Facilities Shutdown Program (IFSP) office was created to execute the stabilization and shutdown plan. The program is entering its third year in which the actual shutdown of the facilities is initiated. Accomplishments to date have included consolidation of all isotopes inventory into one facility, DOE approval of the IFSP Environmental Assessment (EA), and implementation of a detailed management plan for the shutdown of the facilities

Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

Energy Technology Data Exchange (ETDEWEB)

Groth, Katrina, E-mail: kgroth@umd.ed [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States); Wang Chengdong; Mosleh, Ali [Center for Risk and Reliability, 0151 Glenn L. Martin Hall, University of Maryland, College Park, MD 20742 (United States)

2010-12-15

This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

Hybrid causal methodology and software platform for probabilistic risk assessment and safety monitoring of socio-technical systems

International Nuclear Information System (INIS)

Groth, Katrina; Wang Chengdong; Mosleh, Ali

2010-01-01

This paper introduces an integrated framework and software platform for probabilistic risk assessment (PRA) and safety monitoring of complex socio-technical systems. An overview of the three-layer hybrid causal logic (HCL) modeling approach and corresponding algorithms, implemented in the Trilith software platform, are provided. The HCL approach enhances typical PRA methods by quantitatively including the influence of soft causal factors introduced by human and organizational aspects of a system. The framework allows different modeling techniques to be used for different aspects of the socio-technical system. The HCL approach combines the power of traditional event sequence diagram (ESD)event tree (ET) and fault tree (FT) techniques for modeling deterministic causal paths, with the flexibility of Bayesian belief networks for modeling non-deterministic cause-effect relationships among system elements (suitable for modeling human and organizational influences). Trilith enables analysts to construct HCL models and perform quantitative risk assessment and management of complex systems. The risk management capabilities included are HCL-based risk importance measures, hazard identification and ranking, precursor analysis, safety indicator monitoring, and root cause analysis. This paper describes the capabilities of the Trilith platform and power of the HCL algorithm by use of example risk models for a type of aviation accident (aircraft taking off from the wrong runway).

Shutdown cooling temperature perturbation test for analysis of potential flow blockages

International Nuclear Information System (INIS)

Handbury, J.; Newman, C.; Shynot, T.

1996-01-01

This paper details the methods and results of the 'shutdown cooling test' in October 1995. This novel test was conducted at PLGS while the reactor was shutdown and shutdown cooling (SDC) waster was recirculating to find potential channel blockages resulting from the introduction of wood debris. This test discovered most of the channels that contained major wood and metal debris. (author)

Shutdown Safety in NEK

International Nuclear Information System (INIS)

Gluhak, Mario; Senegovic, Marko

2014-01-01

Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity control, continuity of core cooling, and integrity of fission product barriers are valued as essential, distinguishing attributes of nuclear station work environment'. NEK has recognized the importance of 'defence in depth'Industry performance analysis since 2004 has revealed that 23% of the events reported to WANO occurred during outage periods. Given the fact that a plant is in the outage only 5 percent of the time, this emphasizes the importance of shutdown safety and measures station staffs undertake to maintain effective barriers to safety margins during the outage. Back in 1990s, the industry adopted guidance to meet safety requirements by focusing on safety functions. Both WANO and INPO released various documents, reports and guidelines to help accomplish those requirements. However, in the last decade inadequate 'defence in depth' has led to several events affecting shutdown safety and challenging one of the most important nuclear safety principles: 'The special characteristics of nuclear technology are taken into account in all decisions and actions. Reactivity

Fluid shut-down system for a nuclear reactor

International Nuclear Information System (INIS)

Barclay, F.W.; Frey, J.R.; Wilson, J.N.; Besant, R.W.

1975-01-01

A nuclear reactor shut-down system is described which comprises a fluidic vortex valve for releasably maintaining a liquid neutron poison outside of the reactor core, the poison being contained by a reservoir and biased by pressure for flow into poison tubes within the reactor. The upper ends of the poison tubes communicate with the supply port of the vortex valve. A continuous gas flow into the control port maintains normal controlled operation. Shut-down is effected by interruption of the control input. One embodiment comprises three groups of poison tubes and one vortex valve associated with each group wherein shut-down is effected by poison release in two out of the three groups. Preferably, each vortex valve comprises three control ports which operate on a ''voting'' or two-out-of-three basis. (Official Gazette)

Reactor shutdown device

International Nuclear Information System (INIS)

Matsumiya, Hirohito; Endo, Hiroshi; Tsuboi, Yasushi.

1993-01-01

The present invention concerns a reactor shutdown device capable of suppressing change of a core insertion amount relative to temperature change during normal operation and having a great extension amount due to thermal expansion and high mechanical strength. A control rod main body is contained vertically movably in a guide tube disposed in a reactor core. An extension member extends upward from the upper end of a control rod main body and suspends the control rod main body. A shrinkable member intervenes at a midway of the extension member and is made shrinkable. A temperature sensitive member contains coolants at the inside and surrounds the shrinkable member. Thus, if the temperature of external coolants rises abruptly, the shrinkable member is extended by thermal expansion of the coolants in the temperature sensitive member. Upon usual reactor startup, the coolants in the temperature sensitive member cause no substantial thermal expansion by temperature elevation from a cold shutdown temperature to a rated power operation temperature, and the shrinkable member maintains its original state, so that the control rod main body is not inserted into the reactor core. However, upon abrupt temperature elevation, the control rod main body is inserted into the reactor core. (I.S.)

Industry shutdown rates and permanent layoffs: evidence from firm-worker matched data

Directory of Open Access Journals (Sweden)

Kim P. Huynh

2017-06-01

Full Text Available Abstract Firm shutdown creates a turbulent situation for workers as it leads directly to layoffs for its workers. An additional consideration is whether a firm’s shutdown within an industry creates turbulence for workers at other continuing firms. Using data drawn from the Longitudinal Worker File, a Canadian firm-worker matched employment database, we investigate the impact of industry shutdown rates on workers at continuing firm. This paper exploits variation in shutdown rates across industries and within an industry over time to explain the rate of permanent layoffs and the growth of workers’ earnings. We find an increase in industry shutdown rates increases the probability of permanent layoffs and decreases earnings growth for workers at continuing firms.

Application of determination of PRA, Ang II and IGF-1 levels in the study of typing of essential hypertension

International Nuclear Information System (INIS)

Lu Yongyi; Chen Qun; Yang Yongqing

2010-01-01

Objective: To study the clinical application of determination of plasma renin activity (PRA), Angiotensin II (Ang II ) and insulin-like growth factor-1 (IGF-1) levels in typing of essential hypertension (EH). Methods: Determined the levels of PRA and Aug II in 256 patients with EH and 70 healthy volunteers (as control group) by radioimmunoassay, and measured IGF-1 level by enzyme immunoassay. Research on the typing of EH and the difference between the groups. Results: The PRA and Ang II in control group was (0.432±0.236) μg·L -1 ·h -1 and (31.7±7.4) μg/L respectively. In 256 patients with EH, PRA was increased, normal and decreased in 18.0%, 71.8% and 10.2% respectively, while the level of Ang II was increased, normal and decreased in 12.9%, 76.2% and 10.9% respectively. The IGF-1 levels in 256 patients with EH were increased following the increase of blood pressure. Conclusion: Typing of EH patients with PRA and Ang II as well as the determination of IGF-1 were useful in treating and following up the patients with EH. (authors)

Startup and shutdown of the PULSAR Tokamak Reactor

International Nuclear Information System (INIS)

Werley, K.A.; Bathke, C.G.

1994-01-01

Start-up conditions are examined for a pulsed tokamak reactor that uses only inductive plasma current drive for startup, burn and shutdown. A zero-dimensional (profile-averaged) model that describes plasma power and particle balance equations is used to study several aspects of plasma startup and shutdown, including optimization of the startup pathway tradeoff of auxiliary startup heating power versus startup time, volt-second consumtion, thermal stability and partial-power operations

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J.; Best, Ralph E.; Ross, Steven B.; Buxton, Kenneth A.; England, Jeffery L.; McConnell, Paul E.

2013-09-30

This report fulfills the M2 milestone M2FT-13PN0912022, “Stranded Sites De-Inventorying Report.” In January 2013, the U.S. Department of Energy (DOE) issued the Strategy for the Management and Disposal of Used Nuclear Fuel and High-Level Radioactive Waste (DOE 2013). Among the elements contained in this strategy is an initial focus on accepting used nuclear fuel from shutdown reactor sites. This focus is consistent with the recommendations of the Blue Ribbon Commission on America’s Nuclear Future, which identified removal of stranded used nuclear fuel at shutdown sites as a priority so that these sites may be completely decommissioned and put to other beneficial uses (BRC 2012). Shutdown sites are defined as those commercial nuclear power reactor sites where the nuclear power reactors have been shut down and the site has been decommissioned or is undergoing decommissioning. In this report, a preliminary evaluation of removing used nuclear fuel from 12 shutdown sites was conducted. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, and San Onofre. These sites have no other operating nuclear power reactors at their sites and have also notified the U.S. Nuclear Regulatory Commission that their reactors have permanently ceased power operations and that nuclear fuel has been permanently removed from their reactor vessels. Shutdown reactors at sites having other operating reactors are not included in this evaluation.

A Study on Fire Ignition Frequency of UCN 3 during Shutdown

International Nuclear Information System (INIS)

Kim, Kilyoo; Kang, DaeIl; Jang, Seung-Cheol

2014-01-01

A fire ignition frequency of UCN 3 during shutdown, i.e., during POS 3, 4, 5, 6 was calculated by using the new fire PSA method suggested in NUREG/CR-7114. As the fire ignition frequency during full power is calculated by the fixed ignition source and the transient ignition source, the one during shutdown is also calculated by the fixed and the transient ignition source. Since the fixed ignition source was already verified through the walkdown although the walkdown is for the fixed ignition source during full power, additional walkdown for the one during shutdown is not necessary. In the paper, how the fire ignition frequency of UCN 3 during shutdown was calculated is described. A fire ignition frequency of UCN 3 during shutdown, i.e., during POS 3, 4, 5, 6 was calculated by using the new fire PSA method suggested in NUREG/CR-7114. We make the transient ignition fire frequency of each BIN vary according to the daily work order of each POS

Level-1 seismic probabilistic risk assessment for a PWR plant

International Nuclear Information System (INIS)

Kondo, Keisuke; Nishio, Masahide; Fujimoto, Haruo; Ichitsuka, Akihiro

2014-01-01

In Japan, revised Seismic Design Guidelines for the domestic light water reactors was published on September 19, 2006. These new guidelines have introduced the purpose to confirm that residual risk resulting from earthquake that exceeds the design limit seismic ground motion (Ss) is sufficiently small, based on the probabilistic risk assessment (PRA) method, in addition to conventional deterministic design base methodology. In response to this situation, JNES had been working to improve seismic PRA (SPRA) models for individual domestic light water reactors. In case of PWR in Japan, total of 24 plants were grouped into 11 categories to develop individual SPRA model. The new regulatory rules against the Fukushima dai-ichi nuclear power plants' severe accidents occurred on March 11, 2011, are going to be enforced in July 2013 and utilities are necessary to implement additional safety measures to avoid and mitigate severe accident occurrence due to external events such as earthquake and tsunami, by referring to the results of severe accident study including SPRA. In this paper a SPRA model development for a domestic 3-loop PWR plant as part of the above-mentioned 11 categories is described. We paid special attention to how to categorize initiating events that are specific to seismic phenomena and how to confirm the effect of the simultaneous failure probability calculation model for the multiple components on the result of core damage frequency evaluation. Simultaneous failure probability for multiple components has been evaluated by power multiplier method. Then tentative level-1 seismic probabilistic risk assessment (SPRA) has been performed by the developed SPSA model with seismic hazard and fragility data. The base case was evaluated under the condition with calculated fragility data and conventional power multiplier. The difference in CDF between the case of conventional power multiplier and that of power multiplier=1 (complete dependence) was estimated to be

Safety considerations for research reactors in extended shutdown

International Nuclear Information System (INIS)

2004-01-01

According to the IAEA Research Reactor Database, in the last 20 years, 367 research reactors have been shut down. Of these, 109 have undergone decommissioning and the rest are in extended shutdown with no clear definition about their future. Still other research reactors are infrequently operated with no meaningful utilization programme. These two situations present concerns related to safety such as loss of corporate memory, personnel qualification, maintenance of components and systems and preparation and maintenance of documentation. There are many reasons to shut down a reactor; these may include: - the need to carry out modifications in the reactor systems; - the need for refurbishment to extend the lifetime of the reactor; - the need to repair reactor structures, systems, or components; - the need to remedy technical problems; - regulatory or public concerns; - local conflicts or wars; - political convenience; - the lack of resources. While any one of these reasons may lead to shutdown of a reactor, each will present unique problems to the reactor management. The large variations from one research reactor to the next also will contribute to the uniqueness of the problems. Any option that the reactor management adopts will affect the future of the facility. Options may include dealing with the cause of the shutdown and returning to normal operation, extending the shutdown period waiting a future decision, or decommissioning. Such options are carefully and properly analysed to ensure that the solution selected is the best in terms of reactor type and size, period of shutdown and legal, economic and social considerations. This publication provides information in support of the IAEA safety standards for research reactors

BWR startup and shutdown activity transport control

Energy Technology Data Exchange (ETDEWEB)

Garcia, S.E., E-mail: sgarcia@epri.com [Electric Power Research Inst. (EPRI), Palo Alto, California (United States); Giannelli, J.F.; Jarvis, A.J., E-mail: jgiannelli@finetech.com, E-mail: ajarvis@finetech.com [Finetech, Inc., Parsippany, New Jersey (United States)

2010-07-01

This paper summarizes BWR industry experience on good practices for controlling the transport of corrosion product activity during shutdowns, particularly refueling outages, and for startup chemistry control to minimize IGSCC (intergranular stress corrosion cracking). For shutdown, overall goals are to minimize adverse impacts of crud bursts and the time required to remove activated corrosion products from the reactor coolant during the shutdown process prior to refueling, and to assist plants in predicting and controlling radiation exposure during outages. For startup, the overall goals are to highlight conditions during early heatup and startup when sources of reactor coolant oxidants are high, when there is a greater likelihood for chemical excursions associated with refueling outage work activities, and when hydrogen injection is not available to mitigate IGSCC due to system design limitations. BWR water chemistry has changed significantly in recent years with the adoption of hydrogen water chemistry, zinc addition and noble metal chemical applications. These processes have, in some instances, resulted in significant activity increases during shutdown evolutions, which together with reduced time for cleanup because of shorter outages, has consequently increased outage radiation exposure. A review several recent outages shows that adverse effects from these conditions can be minimized, leading to the set of good practice recommendations for shutdown chemistry control. Most plants lose the majority of their hydrogen availability hours during early startup because feedwater hydrogen injection systems were not originally designed to inject hydrogen below 20% power. Hydrogen availability has improved through modifications to inject hydrogen at lower power levels, some near 5%. However, data indicate that IGSCC is accelerated during early startup, when dissolved oxygen and hydrogen peroxide levels are high and reactor coolant temperatures are in the 300 to 400 {sup o

Operating and maintenance experience of Dhruva secondary shutdown system

International Nuclear Information System (INIS)

Sharma, U.L.; Bharathan, R.

1997-01-01

Nine numbers of cadmium shut-off rods are used as primary fast acting shutdown devices while moderator dumping is used as secondary shutdown system. The secondary shutdown system in Dhruva reactor comprises of 3 dump valves and 3 control valves. Under normal operations, the control valves are used to control the moderator level and thereby the reactor power. Under Trip conditions the dump valves as well as the control valves open fully, dumping the moderator to the dump tank, thereby acting as secondary shutdown devices. While the failure of any of these valves to close fully is an incident, the failure of any of these valves to open on a demand is a safety related unusual occurrence and needs to be viewed seriously. During the last 11 years of operation of these valves, there was one incidence of a valve not closing fully and there were two instances of a valve not opening fully on demand. The possible causes, the corrective action taken to rehabilitate these valves and the elaborate system preparations undertaken to enable maintenance jobs are described. (author)

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

International Nuclear Information System (INIS)

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

1994-08-01

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

Shutdown chemistry optimization at Maanshan NPP

International Nuclear Information System (INIS)

Sun Yuanlung; Chuang Benjamin; Su Kouhwa; Kao Jueiting

2009-01-01

At Maanshan PWRs, a significant piping radiation buildup caused by crud burst from fuel surface in the beginning of RFO used to be blamed as a contribution to high personal exposures during outage. Therefore, several modifications on shutdown chemistry procedures such as, early lithium removal, rapid boration, dissolved hydrogen removal, extended RCP operation, and maintaining maximum let down flow, have been consecutively conducted since no.1RFO-16, 2006. The important operational and chemical parameters of modified shutdown chemistry procedures adopted in no.2 RFO-17, 2008 and superiority in low reading (2 mSv/hr) from let down heat exchangers area radiation monitor over 11mSv/hr of no.1 RFO-16 at the same area will be addressed in this paper. At the end of no.2 RFO-17, low personal exposures of 765 man-mSv (TLD)verified the absence of crud burst during shutdown chemistry process and broke records of Maanshan NPP as well. Even with a new job on PZR pre-emptive dissimilar weld overlay which exhausting 17.37% of total 797 man-mSv(TLD) in the latest no.1 RFO-18, 659 man-mSv (TLD) made another record low in the history of Maanshan. (author)

Letter report seismic shutdown system failure mode and effect analysis

International Nuclear Information System (INIS)

KECK, R.D.

1999-01-01

The Supply Ventilation System Seismic Shutdown ensures that the 234-52 building supply fans, the dry air process fans and vertical development calciner are shutdown following a seismic event. This evaluates the failure modes and determines the effects of the failure modes

The roles of NRC research in risk-informed, performance-based regulation

International Nuclear Information System (INIS)

Morrison, D.L.; Murphy, J.A.; Hodges, M.W.; Cunningham, M.A.; Drouin, M.T.; Ramey-Smith, A.M.; VanderMolen, H.

1997-01-01

The NRC is expanding the use of probabilistic risk analysis (PRA) throughout the spectrum of its regulatory activities. The NRC's research program in PRA supports this expansion in a number of ways, from performing basic research to developing guidance for regulatory applications. The author provides an overview of the NRC's PRA research program, then focuses on two key activities - the review of individual plant examinations, and the development of guidance for use of PRA in reactor regulation

Preliminary Evaluation of Removing Used Nuclear Fuel from Shutdown Sites

Energy Technology Data Exchange (ETDEWEB)

Maheras, Steven J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Best, Ralph E. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Ross, Steven B. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Buxton, Kenneth A. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States); England, Jeffery L. [Savannah River Site (SRS), Aiken, SC (United States). Savannah River National Lab. (SRNL); McConnell, Paul E. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Massaro, Lawrence M. [Fermi Research Alliance (FRA), Batavia, IL (United States); Jensen, Philip J. [Pacific Northwest National Lab. (PNNL), Richland, WA (United States)

2016-09-30

A preliminary evaluation of removing spent nuclear fuel (SNF) from 13 shutdown nuclear power plant sites was performed. At these shutdown sites the nuclear power reactors have been permanently shut down and the sites have been decommissioned or are undergoing decommissioning. The shutdown sites were Maine Yankee, Yankee Rowe, Connecticut Yankee, Humboldt Bay, Big Rock Point, Rancho Seco, Trojan, La Crosse, Zion, Crystal River, Kewaunee, San Onofre, and Vermont Yankee. The evaluation was divided into four components: Characterization of the SNF and greater-than-Class C low-level radioactive waste (GTCC waste) inventory A description of the on-site infrastructure at the shutdown sites An evaluation of the near-site transportation infrastructure and transportation experience at the shutdown sites An evaluation of the actions necessary to prepare for and remove SNF and GTCC waste. The primary sources for the inventory of SNF and GTCC waste were the U.S. Department of Energy (DOE) spent nuclear fuel inventory database, industry publications such as StoreFUEL, and government sources such as the U.S. Nuclear Regulatory Commission. The primary sources for information on the conditions of on-site infrastructure and near-site transportation infrastructure and experience included information collected during site visits, information provided by managers at the shutdown sites, Facility Interface Data Sheets compiled for DOE in 2005, Services Planning Documents prepared for DOE in 1993 and 1994, industry publications such as Radwaste Solutions, and Google Earth. State staff, State Regional Group representatives, a Tribal representative, and a Federal Railroad Administration representative have participated in nine of the shutdown site visits. Every shutdown site was found to have at least one off-site transportation mode option for removing its SNF and GTCC waste; some have multiple options. Experience removing large components during reactor decommissioning provided an

Inspection maintenance and planning of shutdown in thermal electric generating plants

International Nuclear Information System (INIS)

Dezordi, W.L.; Correa, D.A.; Kina, M.

1984-01-01

The schedule shutdown of an industrial plant and, more specifically, of an electrical generating station, is becoming increasingly important. The major parameters to be taken into account for the planning of such a shutdown are basically of economic-financial nature such as costs of the related services (materials, equipment, manpower, etc), loss of revenue caused by the station's shutdown as well as by the station availability, and other requirements expected from it by the Load Dispatch and consumers. Improving the equipment's performances and the station's availability are the fundamental objectives to be strived for. The authors present in this paper, in an abridged form, the planning tools used for thermal electric generating plants shutdowns for inspections, maintenance and design changes implementation. (Author) [pt

77 FR 75198 - Standard Format and Content for Post-Shutdown Decommissioning Activities Report

Science.gov (United States)

2012-12-19

... NUCLEAR REGULATORY COMMISSION [NRC-2012-0299] Standard Format and Content for Post-Shutdown... regulatory guide (DG), DG-1272, ``Standard Format and Content for Post-shutdown Decommissioning Activities... Content for Post-shutdown Decommissioning Activities Report,'' which was issued in July 2000. DG-1271...

40 CFR 63.310 - Requirements for startups, shutdowns, and malfunctions.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 9 2010-07-01 2010-07-01 false Requirements for startups, shutdowns, and malfunctions. 63.310 Section 63.310 Protection of Environment ENVIRONMENTAL PROTECTION AGENCY... CATEGORIES National Emission Standards for Coke Oven Batteries § 63.310 Requirements for startups, shutdowns...

Shutdowns/scrams at BWRs reported under new 1984 LER rule

International Nuclear Information System (INIS)

Mays, G.T.

1985-01-01

Operating experience data from nuclear power plants are essential for safety and reliability analyses. The Licensee Event Reports (LERs), submitted to the NRC by nuclear power plant utilities, contain much of this data. One of the significant aspects of the new LER rule includes the requirement to report all plant shutdowns whereas prior to 1984, not all shutdowns were reported as LERs. This paper reviews the shutdowns and scrams occurring during the first six months of 1984 at BWRs as reported under the new LER rule. The review focused on systems involved, causes, and personnel interactions

Alternative Shutdown Panel. Amaraz Nuclear Power Plant

Energy Technology Data Exchange (ETDEWEB)

Saenz de Santa Maria Valin, J.

2016-07-01

Between 2010 and 2014 the Nuclear Power Plant of Almaraz conducted one of the most complex projects in its history: The installation of an Alternative Shutdown Panel with the capability to stop the plant in case of fire in the Control room or in the Cable room. This project represented a great economic and organizational effort for the plant, but at the same time has been a great improvement in the safety of the installation, which was demonstrated by the achievement of a major milestone in the history of Almaraz: The actual shutdown from outside of the Control room. (Author)

Training simulator for advanced gas-cooled reactor (AGR) shutdown sequence equipment

International Nuclear Information System (INIS)

Shankland, J.P.; Nixon, G.L.

1978-01-01

Successful shutdown of nuclear plant is of prime importance for both safety and economic reasons and large sums of money are spent on equipment to make shutdowns fully automatic, thus removing the possibility of operator errors. While this aim can largely be realized, one must consider the possibility of automatic equipment or plant failures when operators are required to take manual action, and off-line training facilities should be available to operating staff to minimize the risk of incorrect actions being taken. This paper presents the practice adopted at Hunterston 'B' Nuclear Power Station to solve this problem and concerns the computer-based training simulator for the Reactor Shutdown Sequence Equipment (RSSE) which was commissioned in January 1977. The plant associated with shutdown is briefly described and the reasoning which shows the need for a simulator is outlined. The paper also gives details of the comprehensive facilities available on the simulator and goes on to describe the form that shutdown training takes and the experience gained at this time. (author)

Development of self-actuated shutdown system using curie point electromagnet

International Nuclear Information System (INIS)

Kim, Tae Ryong; Park, Jin Ho

1999-01-01

An innovative concept for a passive reactor shutdown system, so called self-actuated shutdown system (SASS), is inevitably required for the inherent safety in liquid metal reactor, which is designed with the totally different concept from the usual reactor shutdown system in LWR. SASS using Curie point electromagnet (CPEM) was selected as the passive reactor shutdown system for KALIMER (Korea Advanced Liquid Metal Reactor). A mock-up of the SASS was designed, fabricated and tested. From the test it was confirmed that the mockup was self-actuated at the Curie point of the temperature sensing material used in the mockup. An articulated control rod was also fabricated and assembled with the CPEM to confirm that the control rod can be inserted into core even when the control rod guide tube is deformed due to earthquake. The operability of SASS in the actual sodium environment should be confirmed in the future. All the design and test data will be applied to the KALIMER design. (author)

Recommendations for a proposed standard for performing systems analysis

International Nuclear Information System (INIS)

LaChance, J.; Whitehead, D.; Drouin, M.

1998-01-01

In August 1995, the Nuclear Regulatory Commission (NRC) issued a policy statement proposing improved regulatory decisionmaking by increasing the use of PRA [probabilistic risk assessment] in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data. A key aspect in using PRA in risk-informed regulatory activities is establishing the appropriate scope and attributes of the PRA. In this regard, ASME decided to develop a consensus PRA Standard. The objective is to develop a PRA Standard such that the technical quality of nuclear plant PRAs will be sufficient to support risk-informed regulatory applications. This paper presents examples recommendations for the systems analysis element of a PRA for incorporation into the ASME PRA Standard

Human reliability analysis methods for probabilistic safety assessment

International Nuclear Information System (INIS)

Pyy, P.

2000-11-01

Human reliability analysis (HRA) of a probabilistic safety assessment (PSA) includes identifying human actions from safety point of view, modelling the most important of them in PSA models, and assessing their probabilities. As manifested by many incidents and studies, human actions may have both positive and negative effect on safety and economy. Human reliability analysis is one of the areas of probabilistic safety assessment (PSA) that has direct applications outside the nuclear industry. The thesis focuses upon developments in human reliability analysis methods and data. The aim is to support PSA by extending the applicability of HRA. The thesis consists of six publications and a summary. The summary includes general considerations and a discussion about human actions in the nuclear power plant (NPP) environment. A condensed discussion about the results of the attached publications is then given, including new development in methods and data. At the end of the summary part, the contribution of the publications to good practice in HRA is presented. In the publications, studies based on the collection of data on maintenance-related failures, simulator runs and expert judgement are presented in order to extend the human reliability analysis database. Furthermore, methodological frameworks are presented to perform a comprehensive HRA, including shutdown conditions, to study reliability of decision making, and to study the effects of wrong human actions. In the last publication, an interdisciplinary approach to analysing human decision making is presented. The publications also include practical applications of the presented methodological frameworks. (orig.)

The NUREG-1150 probabilistic risk assessment for the Grand Gulf nuclear station

International Nuclear Information System (INIS)

Brown, T.D.; Breeding, R.J.; Jow, H.N.; Higgins, S.J.; Shiver, A.W.; Helton, J.C.

1992-01-01

This paper summarizes the findings of the probabilistic risk assessment (PRA) for Unit 1 of the Grand Gulf Nuclear Station performed in support of NUREG-1150. The emphasis is on the 'back-end' analyses, that is, the acident progression, source term, consequence analsyes, and risk results obtained when the results of these analyses are combined with the accident frequency analysis. The offsite risk from internal initiating events was found to be quite low, both with respect to the safety goals and to the other plants analyzed in NUREG-1150. The offsite risk is dominated by short-term station blackout plant damage states. The long-term blackout group and the anticiptated transients without scram (ATWS) group contribute considerably less to risk. Transients in which the power conversion system is unavailable are very minor contributors to risk. The low values for risk can be attributed to low core damage frequency, good emergency response, and plant features that reduce the potential source term. (orig.)

30 CFR 57.8534 - Shutdown or failure of auxiliary fans.

Science.gov (United States)

2010-07-01

... 30 Mineral Resources 1 2010-07-01 2010-07-01 false Shutdown or failure of auxiliary fans. 57.8534... Ventilation Underground Only § 57.8534 Shutdown or failure of auxiliary fans. (a) Auxiliary fans installed and... fan maintenance or fan adjustments where air quality is maintained in compliance with the applicable...

Probabilistic safety analysis of DC power supply requirements for nuclear power plants. Technical report

International Nuclear Information System (INIS)

Baranowsky, P.W.; Kolaczkowski, A.M.; Fedele, M.A.

1981-04-01

A probabilistic safety assessment was performed as part of the Nuclear Regulatory Commission generic safety task A-30, Adequacy of Safety Related DC Power Supplies. Event and fault tree analysis techniques were used to determine the relative contribution of DC power related accident sequences to the total core damage probability due to shutdown cooling failures. It was found that a potentially large DC power contribution could be substantially reduced by augmenting the minimum design and operational requirements. Recommendations included (1) requiring DC power divisional independence, (2) improved test, maintenance, and surveillance, and (3) requiring core cooling capability be maintained following the loss of one DC power bus and a single failure in another system

PSA for the shutdown mode for nuclear power plants

International Nuclear Information System (INIS)

1994-06-01

The meeting, which was attended by more than 75 participants from 20 countries, provided a broad discussion forum where all the currently active major shutdown PSA programmes were reviewed. The meeting also addressed the issues related to actual performance of shutdown PSA studies as well as insight gained from the studies. This document, which was prepared during the TCM, contains the results of extensive discussions which were held in specific working groups. The papers presented at the meeting provide a comprehensive overview of the state of the art of shutdown risk assessment and remedial measures taken to reduce the risk in outages. It is hoped that this document will be very useful to all individuals with interest in increasing safety during outages at NPPs. Refs, figs and tabs

Core shutdown report: Subcycle K-14.1

International Nuclear Information System (INIS)

Gough, S.T.

1992-05-01

When a reactor is shut down, there is a set of rules that must be followed to guarantee that the reactor remains in a safe shutdown state. Some of these rules involve the cooling of heat generating assemblies before, during, and after charge-discharge (C ampersand D) operations. These rules ensure that C ampersand D operations will not endanger the integrity of the fuel or targets by allowing them to overheat. DPSOL 105-1225, Assembly Discharge and Forced Cooling Requirements, is the primary operations procedure that governs these cooling rules. The specific shutdown cooling limits that are input into this procedure are contained within this report

Improvements of primary coolant shutdown chemistry and reactor coolant system cleanup

International Nuclear Information System (INIS)

Gaudard, G.; Gilles, B.; Mesnage, F.; Cattant, F.

2002-01-01

In the framework of a radiation exposure management program entitled >, EDF aims at decreasing the mass dosimetry of nuclear power plants workers. So, the annual dose per unit, which has improved from 2.44 m.Sv in 1991 to 1.08 in 2000, should target 0.8 mSv in the year 2005 term in order to meet the results of the best nuclear operators. One of the guidelines for irradiation source term reduction is the optimization of operation parameters, including reactor coolant system (RCS) chemistry in operation, RCS shutdown chemistry and RCS cleanup improvement. This paper presents the EDF strategy for the shutdown and start up RCS chemistry optimization. All the shutdown modes have been reviewed and for each of them, the chemical specifications will be fine tuned. A survey of some US PWRs shutdown practices has been conducted for an acid and reducing shutdown chemistry implementation test at one EDF unit. This survey shows that deviating from the EPRI recommended practice for acid and reducing shutdown chemistry is possible and that critical path impact can be minimized. The paper also presents some investigations about soluble and insoluble species behavior and characterization; the study focuses here on 110m Ag, 122 Sb, 124 Sb and iodine contamination. Concerning RCS cleanup improvement, the paper presents two studies. The first one highlights some limited design modifications that are either underway or planned, for an increased flow rate during the most critical periods of the shutdown. The second one focuses on the strategy EDF envisions for filters and resins selection criteria. Matching the study on contaminants behavior with the study of filters and resins selection criteria should allow improving the cleanup efficiency. (authors)

Probabilistic Decision Graphs - Combining Verification and AI Techniques for Probabilistic Inference

DEFF Research Database (Denmark)

Jaeger, Manfred

2004-01-01

We adopt probabilistic decision graphs developed in the field of automated verification as a tool for probabilistic model representation and inference. We show that probabilistic inference has linear time complexity in the size of the probabilistic decision graph, that the smallest probabilistic ...

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

Energy Technology Data Exchange (ETDEWEB)

C. L. Smith; K. J. Kvarfordt; S. T. Wood

2008-08-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

Energy Technology Data Exchange (ETDEWEB)

C. L. Smith; K. J. Kvarfordt; S. T. Wood

2006-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for ansforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with

Learning Probabilistic Logic Models from Probabilistic Examples.

Science.gov (United States)

Chen, Jianzhong; Muggleton, Stephen; Santos, José

2008-10-01

We revisit an application developed originally using abductive Inductive Logic Programming (ILP) for modeling inhibition in metabolic networks. The example data was derived from studies of the effects of toxins on rats using Nuclear Magnetic Resonance (NMR) time-trace analysis of their biofluids together with background knowledge representing a subset of the Kyoto Encyclopedia of Genes and Genomes (KEGG). We now apply two Probabilistic ILP (PILP) approaches - abductive Stochastic Logic Programs (SLPs) and PRogramming In Statistical modeling (PRISM) to the application. Both approaches support abductive learning and probability predictions. Abductive SLPs are a PILP framework that provides possible worlds semantics to SLPs through abduction. Instead of learning logic models from non-probabilistic examples as done in ILP, the PILP approach applied in this paper is based on a general technique for introducing probability labels within a standard scientific experimental setting involving control and treated data. Our results demonstrate that the PILP approach provides a way of learning probabilistic logic models from probabilistic examples, and the PILP models learned from probabilistic examples lead to a significant decrease in error accompanied by improved insight from the learned results compared with the PILP models learned from non-probabilistic examples.

Is probability of frequency too narrow?

International Nuclear Information System (INIS)

Martz, H.F.

1993-01-01

Modern methods of statistical data analysis, such as empirical and hierarchical Bayesian methods, should find increasing use in future Probabilistic Risk Assessment (PRA) applications. In addition, there will be a more formalized use of expert judgment in future PRAs. These methods require an extension of the probabilistic framework of PRA, in particular, the popular notion of probability of frequency, to consideration of frequency of frequency, frequency of probability, and probability of probability. The genesis, interpretation, and examples of these three extended notions are discussed

The Alternative Design Features for Safety Enhancement in Shutdown Operation

International Nuclear Information System (INIS)

Oh, Hae Cheol; Kim, Myung Ki; Chung, Bag Soon; Seo, Mi Ro

2009-01-01

PSA can be used to confirm that the new plant design is complied with the applicable safety goals, and to select among the alternate design options. A shutdown PSA provides insight for outage planning schedule, outage management practices, and design modifications. Considering the results of both LPSD PSA studies and operating experiences for low power and shutdown, the improvements can be proposed to reduce the high risk contribution. The improvements/enhancements during shutdown operation may be divided into categories such as hardware, administrative management, and operational procedure. This paper presents on an example how the risk related to an accidental situation can be reduced, focusing the hardware design changes for the newly designed NPPs

Study of methodology for low power/shutdown fire PSA

International Nuclear Information System (INIS)

Yan Zhen; Li Zhaohua; Li Lin; Song Lei

2014-01-01

As a risk assessment technology based on probability, the fire PSA is accepted abroad by nuclear industry in its application in the risk assessment for nuclear power plants. Based on the industry experience, the fire-induced impact on the plant safety during low power and shutdown operation cannot be neglected, therefore fire PSA can be used to assess the corresponding fire risk. However, there is no corresponding domestic guidance/standard as well as accepted analysis methodology up to date. Through investigating the latest evolvement on fire PSA during low power and shutdown operation, and integrating its characteristic with the corresponding engineering experience, an engineering methodology to evaluate the fire risk during low power and shutdown operation for nuclear power plant is established in this paper. In addition, an analysis demonstration as an example is given. (authors)

Updating of the program for simulation of Darlington shutdown and regulation systems

International Nuclear Information System (INIS)

1988-07-01

This report describes the current status of the developments of a simulation of the Darlington Nuclear Generating Station shutdown and regulating systems, DARSIM done under contract to the Atomic Energy Control Board (AECB). The DARSIM program simulates the spatial neutron dynamics, the regulation of the reactor power, and shutdown system 1 and shutdown system 2 software. The DARSIM program operates in the interactive simulation program environment. DARSIM was installed on the APOLLO computer at the AECB and a version for an IBM-PC was also provided for the exclusive use of the AECB. Shutdown system software was updated to incorporate the latest revisions in the functional specifications. Additional developments have been provided to assist in the use and interpretation of the DARSIM results

Backup passive reactivity shutdown systems

International Nuclear Information System (INIS)

Ashurko, Yu.M.; Kuznetsov, L.A.

1996-01-01

The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs

Backup passive reactivity shutdown systems

Energy Technology Data Exchange (ETDEWEB)

Ashurko, Yu M; Kuznetsov, L A [Institute of Physics and Power Engineering, Obninsk (Russian Federation)

1996-12-01

The paper reviews self-actuated shutdown systems (SASSs) for liquid metal-cooled fast reactors (LMFRs). Principles of operation are described, advantages and drawbacks analyzed, and prospects for application in advanced fast reactors examined. Ways to improve reactor self-protection via reactivity feedback amplification and related problems are discussed. (author). 9 refs, 12 figs.

Comparison of SKIFS 2004:1 and Tillsynshandbok PSA against the ASME PRA Standard and European requirements on PSA

International Nuclear Information System (INIS)

Hellstroem, Per

2005-04-01

Requirements on PSA for risk informed applications are expressed in different international documents. The ASME PRA standard published in spring 2002 is one such document, PSA requirements are also expressed in the European Utility Requirements (EUR) for new reactors. The Swedish PSA requirements are provided in the Swedish regulators (SKI) statutes SKIFS 2004:1. SKI also has a review handbook for PSA activities (SKI report 2003:48). The review handbook is a support during review of the utilities PSA activities and the PSAs themselves. The review handbook expresses SKIs expectations by providing so called important aspects for both the PSA work and the PSAs, A comparison of SKIFS requirements and the important aspects in the Review handbook, on one side, and the requirements on PSA in EUR and ASME on the other side, is presented. The comparison shows a large difference in the level of detail in the different documents, where ASME is most detailed and specific. This is expected since the SKI review handbook not is a 'PSA guide' in the same way as the ASME PRA standard. A direct comparison of the ASME PRA standard requirements with the important aspects in the review handbook cannot answer the question which ASME capacity level that is achieved by a PSA meeting all important aspects. The conclusion is that it is not likely to achieve capacity level 2 and 3, since very few ASME level 3 attributes are explicitly expressed as important aspects, though many are expressed in general terms. The review handbook important aspects that are most similar to the ASME capacity level 1 attributes are initiating events, sequence analysis, and system analysis while less similarity is found for analysis of operator actions data analysis, quantification and containment analysis (level 2). Less similarity is found for capacity level 2 and 3. However, the number of additional ASME attributes on capacity level 2 and 3 are few. There are also important aspects in the review handbook that

Examination of risk significant configuration during low power and shutdown with ORION and PSA

Energy Technology Data Exchange (ETDEWEB)

Park, Chul Kyu; Oh, Seung Jong [KEPCO International Nuclear Graduate School, Ulsan (Korea, Republic of)

2015-10-15

This paper suggests an approach to calculate the increased CDF corresponding to Orange and Red states in ORION program and analyzed the result of calculation. This approach is expected to be useful for checking the adequacy of the LPSD PSA. And also, the result of this calculation can provide the information about which SSCs for certain SF are more sensitive to risk in particular POS. Defense-in-depth is a safety philosophy in which multiple lines of defense and conservative design and evaluation methods are applied to ensure the safety of the public. Based on this philosophy EPRI developed Outage Risk Assessment and Management (ORAM) program as a qualitative assessment to better manage the risk during low power and shutdown event after the Vogtle loss of vital AC power and RHR event in 1990. Each risk level of RED, ORANGE color status caused by the degradation of each key safety function might be different depend on the importance of each key safety function. However we can't know how much different. If we know the quantitative information about the risk level represented by color, we can take and prepare concrete actions to reduce the risk level of the plant with rescheduling maintenance, strengthen surveillance for important safety function, and developing outage management strategy. The probabilistic safety analysis for low power and shutdown period can provide risk information with quantitative value related on the degradation of redundancy and diversity level for the safety functions during outage. In this study, we calculated the increased Core Damage frequency (CDF) of each RED and ORANGE states in ORION program caused by the degradation of each key safety function by modifying LPSD PSA model. The result of calculation and analysis could be effective to check adequacy and find improvement for these two methods.

Loss-of-benefits analysis for nuclear power plant shutdowns: methodology and illustrative case study

International Nuclear Information System (INIS)

Peerenboom, J.P.; Buehring, W.A.; Guziel, K.A.

1983-11-01

A framework for loss-of-benefits analysis and a taxomony for identifying and categorizing the effects of nuclear power plant shutdowns or accidents are presented. The framework consists of three fundamental steps: (1) characterizing the shutdown; (2) identifying benefits lost as a result of the shutdown; and (3) quantifying effects. A decision analysis approach to regulatory decision making is presented that explicitly considers the loss of benefits. A case study of a hypothetical reactor shutdown illustrates one key loss of benefits: net replacement energy costs (i.e., change in production costs). Sensitivity studies investigate the responsiveness of case study results to changes in nuclear capacity factor, load growth, fuel price escalation, and discount rate. The effects of multiple reactor shutdowns on production costs are also described

Effect of dc-power-system reliability on reactor-shutdown cooling

International Nuclear Information System (INIS)

Kolaczkowski, A.M.; Baranowsky, P.W.; Hickman, J.W.

1981-01-01

The DC power systems in a nuclear power plant provide control and motive power to valves, instrumentation, emergency diesel generators, and many other components and systems during all phases of plant operation including abnormal shutdowns and accident situations. A specific area of concern is the adequacy of the minimum design requirements for DC power systems, particularly with regard to multiple and common cause failures. This concern relates to the application of the single failure criterion for assuring a reliable DC power supply which may be required for the functionability of shutdown cooling systems. The results are presented of a reliability based study performed to assess the adequacy of DC power supply design requirements for currently operating light water reactors with particular attention to shutdown cooling requirements

Post Fire Safe Shutdown Analysis Using a Fault Tree Logic Model

International Nuclear Information System (INIS)

Yim, Hyun Tae; Park, Jun Hyun

2005-01-01

Every nuclear power plant should have its own fire hazard analysis including the fire safe shutdown analysis. A safe shutdown (SSD) analysis is performed to demonstrate the capability of the plant to safely shut down for a fire in any given area. The basic assumption is that there will be fire damage to all cables and equipment located within a common fire area. When evaluating the SSD capabilities of the plant, based on a review of the systems, equipment and cables within each fire area, it should be determined which shutdown paths are either unaffected or least impacted by a postulated fire within the fire area. Instead of seeking a success path for safe shutdown given all cables and equipment damaged by a fire, there can be an alternative approach to determine the SSD capability: fault tree analysis. This paper introduces the methodology for fire SSD analysis using a fault tree logic model

New probabilistic decision-making tools for fire protection systems

International Nuclear Information System (INIS)

Ksobiech, C.; Mowrer, F.

1991-01-01

The FIVE methodology provides guidance to utilities in performing an examination of potential plant severe accidents caused by fire initiated events. FIVE is oriented toward uncovering limiting plant design or operating characteristics (vulnerabilities) that make certain fire-initiated events more likely than others. It provides a combination of deterministic and probabilistic techniques for examining a power plant's fire probability and protection characteristics. It includes a two phase progressive screening method and a third phase consisting of a plant walkdown/verification process. The FIVE methodology centers on providing assurance of the availability of at least one train of the safe shutdown systems. FIVE has been developed for implementation by plant personnel who are most experienced with their plant's operations, fire hazards and fire protection features. The methodology provides these plant personnel with guidelines to quickly screen the plant down to the most significant locations where vulnerabilities may exist and then identify options to reduce the vulnerabilities

Expected proton signal sizes in the PRaVDA Range Telescope for proton Computed Tomography

International Nuclear Information System (INIS)

Price, T.; Parker, D.J.; Green, S.; Esposito, M.; Waltham, C.; Allinson, N.M.; Poludniowski, G.; Evans, P.; Taylor, J.; Manolopoulos, S.; Anaxagoras, T.; Nieto-Camero, J.

2015-01-01

Proton radiotherapy has demonstrated benefits in the treatment of certain cancers. Accurate measurements of the proton stopping powers in body tissues are required in order to fully optimise the delivery of such treaments. The PRaVDA Consortium is developing a novel, fully solid state device to measure these stopping powers. The PRaVDA Range Telescope (RT), uses a stack of 24 CMOS Active Pixel Sensors (APS) to measure the residual proton energy after the patient. We present here the ability of the CMOS sensors to detect changes in the signal sizes as the proton traverses the RT, compare the results with theory, and discuss the implications of these results on the reconstruction of proton tracks

Recent case studies and advancements in probabilistic risk assessment

International Nuclear Information System (INIS)

Garrick, B.J.

1985-01-01

During the period from 1977 to 1984, Pickard, Lowe and Garrick, Inc., had the lead in preparing several full scope probabilistic risk assessments for electric utilities. Five of those studies are discussed from the point of view of advancements and lessons learned. The objective and trend of these studies is toward utilization of the risk models by the plant owners as risk management tools. Advancements that have been made are in presentation ad documentation of the PRAs, generation of more understandable plant level information, and improvements in methodology to facilitate technology transfer. Specific areas of advancement are in the treatment of such issues as dependent failures, human interaction, and the uncertainty in the source term. Lessons learned cover a wide spectrum and include the importance of plant specific models for meaningful risk management, the role of external events in risk, the sensitivity of contributors to choice of risk index, and the very important finding that the public risk is extremely small. The future direction of PRA is to establish less dependence on experts for in-plant application. Computerizing the PRAs such that they can be accessed on line and interactively is the key

40 CFR 65.6 - Startup, shutdown, and malfunction plan and procedures.

Science.gov (United States)

2010-07-01

... 40 Protection of Environment 15 2010-07-01 2010-07-01 false Startup, shutdown, and malfunction... (CONTINUED) AIR PROGRAMS (CONTINUED) CONSOLIDATED FEDERAL AIR RULE General Provisions § 65.6 Startup... Group 2A or Group 2B process vents. (b) Startup, shutdown, and malfunction plan—(1) Description and...

Maintenance, repair and operation (MRO) of shutdown facilities

International Nuclear Information System (INIS)

Kenny, S.

2006-01-01

What level of maintenance does one apply to a shutdown facility? Well it depends on who you ask. Operations staff sees facilities that have completed their useful life cycle as a cost drain while Decommissioning staff sees this as the start of a new life cycle. Based on the decommissioning plan for the particular facility the building could complete another full life cycle while under decommissioning whether it is in storage with surveillance mode or under active decommissioning. This paper will explore how you maintain a facility and systems for many years after its useful life until final decommissioning is completed. When a building is declared redundant, who looks after it until the final decommissioning end state is achieved? At the AECL, Chalk River Labs site the safe shutdown and turnover process is one key element that initiates the decommissioning process. The real trick is orchestrating maintenance, repair and operation plans for a facility that has been poorly invested in during its last years of useful life cycle. To add to that usually shutdowns are prolonged for many years beyond the expected turnover period. During this presentation I will cover what AECL is doing to ensure that the facilities are maintained in a proper state until final decommissioning can be completed. All facilities or systems travel through the same life cycle, design, construction, commissioning, operation, shutdown and demolition. As we all know, nuclear facilities add one more interesting twist to this life cycle called Decommissioning that lands between shutdown and demolition. As a facility nears the shutdown phase, operations staff loose interest in the facility and stop investing in upgrades, repairs and maintenance but continue to invest and focus on maximizing operations. Facility maintenance standards produced by the International Facility Maintenance Association (IFMA) based on a survey done every year state that 2.2% of the total operating costs for the site should be

Probabilistic Risk Assessment Process for High-Power Laser Operations in Outdoor Environments

Science.gov (United States)

2016-01-01

the NOHD/NSHD or its derivatives. Only those issued with an appropriate level of protection (such as protective eyewear or clothing) would be...developed in the emerging nuclear industry as well as in the established transport, petrochemical, and aerospace sectors, which were expanding...emergence of the PRA technique as an effective means of risk assessment. The origins of PRA lie in the aerospace industry .11,12 PRA is described by the

Reliability of Offshore Wind Turbine Drivetrains based on Measured Shut-down Events

DEFF Research Database (Denmark)

Natarajan, Anand; Buhl, Thomas

2015-01-01

by initiating blade pitching to feather and also sometimes using the generator torqueas a brake mechanism. The shutdowns due to wind speed variation nearcut-out are predicted using an Inverse First Order Reliability Model(IFORM) whereby an expected annual frequency of normal shutdownsat cut-out is put forth...... normal operation and with shutdowns. The maximum coefficient of variation (CoV) due to varying wind conditions was found on the low speed shaft torsion, but the shutdowns by themselves were not seento significantly change the fatigue loads....

Risk-ranking IST components into two categories

International Nuclear Information System (INIS)

Rowley, C.W.

1996-01-01

The ASME has utilized several schemes for identifying the appropriate scope of components for inservice testing (IST). The initial scope was ASME Code Class 1/2/3, with all components treated equally. Later the ASME Operations and Maintenance (O ampersand M) Committee decided to use safe shutdown and accident mitigation as the scoping criteria, but continued to treat all components equal inside that scope. Recently the ASME O ampersand M Committee decided to recognize service condition of the component, hence the comprehensive pump test. Although probabilistic risk assessments (PRAs) are incredibly complex plant models and computer hardware and software intensive, they are a tool that can be utilized by many plant engineering organizations to analyze plant system and component applications. In 1992 the ASME O ampersand M Committee got interested in using the PRA as a tool to categorize its pumps and valves. In 1994 the ASME O ampersand M Committee commissioned the ASME Center for Research and Technology Development (CRTD) to develop a process that adapted the PRA technology to IST. In late 1995 that process was presented to the ASME O ampersand M Committee. The process had three distinct portions: (1) risk-rank the IST components; (2) develop a more effective testing strategy for More Safety Significant Components; and (3) develop a more economic testing strategy for Less Safety Significant Components

Risk-ranking IST components into two categories

Energy Technology Data Exchange (ETDEWEB)

Rowley, C.W.

1996-12-01

The ASME has utilized several schemes for identifying the appropriate scope of components for inservice testing (IST). The initial scope was ASME Code Class 1/2/3, with all components treated equally. Later the ASME Operations and Maintenance (O&M) Committee decided to use safe shutdown and accident mitigation as the scoping criteria, but continued to treat all components equal inside that scope. Recently the ASME O&M Committee decided to recognize service condition of the component, hence the comprehensive pump test. Although probabilistic risk assessments (PRAs) are incredibly complex plant models and computer hardware and software intensive, they are a tool that can be utilized by many plant engineering organizations to analyze plant system and component applications. In 1992 the ASME O&M Committee got interested in using the PRA as a tool to categorize its pumps and valves. In 1994 the ASME O&M Committee commissioned the ASME Center for Research and Technology Development (CRTD) to develop a process that adapted the PRA technology to IST. In late 1995 that process was presented to the ASME O&M Committee. The process had three distinct portions: (1) risk-rank the IST components; (2) develop a more effective testing strategy for More Safety Significant Components; and (3) develop a more economic testing strategy for Less Safety Significant Components.

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C Systems

Energy Technology Data Exchange (ETDEWEB)

Jang, Seung Cheol; Kang, Hyung Gook; Lim, Ho Gon; Park, Jin Hee; Eom, Heung Sub; Kim, Tae Woon; Ha, Jae Joo

2005-04-15

There are two technical areas to deal with in the project; the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: Quality assessment of a LPSD PSA model for a Korean Standard Nuclear Power Plant (KSNP), Quality improvement of the KSNP LPSD PSA model in the following four technical areas; plant operating status (POS), initiating event analysis, determination of success criteria, accident sequence analysis, Development of the LPSD risk management technologies, Unavailability analysis of Digital safety systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS), Impact analysis of the digital safety systems on plant risks throughout of the digital plant risk models for evaluating core damage frequency (CDF) and large early release frequency (LERF), Study on the methodologies for treating digital-specific problems in the digital I and C PSA such as reliability of safety-critical software, common cause failure (CCF) of digital components, fault coverage, etc.

Development of Risk Assessment Technology for Low Power, Shutdown and Digital I and C Systems

International Nuclear Information System (INIS)

Jang, Seung Cheol; Kang, Hyung Gook; Lim, Ho Gon; Park, Jin Hee; Eom, Heung Sub; Kim, Tae Woon; Ha, Jae Joo

2005-04-01

There are two technical areas to deal with in the project; the low power and shutdown probabilistic safety assessment (PSA), and the digital I and C PSA. The scope and contents of each area could be summarized as follows: Quality assessment of a LPSD PSA model for a Korean Standard Nuclear Power Plant (KSNP), Quality improvement of the KSNP LPSD PSA model in the following four technical areas; plant operating status (POS), initiating event analysis, determination of success criteria, accident sequence analysis, Development of the LPSD risk management technologies, Unavailability analysis of Digital safety systems such as Digital Plant Protection System (DPPS) and Digital Engineered Safety Feature Actuation System (DESFAS), Impact analysis of the digital safety systems on plant risks throughout of the digital plant risk models for evaluating core damage frequency (CDF) and large early release frequency (LERF), Study on the methodologies for treating digital-specific problems in the digital I and C PSA such as reliability of safety-critical software, common cause failure (CCF) of digital components, fault coverage, etc

Analysis of solutions for passively activated safety shutdown devices for SFR

International Nuclear Information System (INIS)

Burgazzi, Luciano

2013-01-01

Highlights: • Innovative systems for emergency shut down of fast reactors are proposed. • The concepts of inherent and passive safety are put forward. • The relative analysis in terms of safety and reliability is presented. • A comparative assessment among the concepts is performed. • Path forward is tracked. -- Abstract: In order to enhance the inherent safety of fast reactors, innovative reactivity control systems have been proposed for intrinsic ultimate shut-down instead of conventional scram rods, to cope with the potential consequences of severe unprotected transient accidents, such as an energetic core disruptive accident, as in case of sodium fast reactors. The passive shut-down systems are designed to shut-down system only by inherent passive reactivity feedback mechanism, under unprotected accident conditions, implying failure of reactor protection system. They are conceived to be self-actuated without any signal elaboration, since the actuation of the system is triggered by the effects induced by the transient like material dilatation, in case of overheating of the coolant for instance, according to fast reactor design to meet the safety requirements. This article looks at different special shutdown systems specifically engineered for prevention of severe accidents, to be implemented on fast reactors, with main focus on the investigation of the performance of the self-actuated shutdown systems in sodium fast reactors

40 CFR 62.15150 - What happens to the operating requirements during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... requirements during periods of startup, shutdown, and malfunction? 62.15150 Section 62.15150 Protection of... § 62.15150 What happens to the operating requirements during periods of startup, shutdown, and... municipal waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or...

Nuclear reactor shutdown system

International Nuclear Information System (INIS)

Mangus, J.D.; Cooper, M.H.

1982-01-01

An improved nuclear reactor shutdown system is described comprising a temperature sensitive device connected to control the electric power supply to a magnetic latch holding a body of a neutron absorbing material. The temperature sensitive device is exposed to the reactor coolant so that when the reactor coolant temperature rises above a specific level, the temperature sensitive device will cause deenergization of the magnetic latch to allow the body of neutron absorbing material to enter the reactor core. (author)

Prioritization of motor operated valves based on risk importances

International Nuclear Information System (INIS)

Vesely, W.E.; Weidenhamer, G.H.

1994-01-01

The plant Probabilistic Risk Assessment (PRA) can be a potentially useful and powerful tool for helping to define an effective response to GL 89-10. The plant PRA can be used to prioritize the Motor Operated Valves (MOV) dynamic test. The plant PRA can also be used to determine test schedules for the MOVs. In order for the PRA to be validly used to respond to GL 89-10, various issues need to be validly addressed. Eleven issues are specifically identified and responses to these issues are outlined. The issues of joint MOV importance, PRA truncation, and validation of the proposed approach are specifically highlighted and more detailed response considerations are described. As in all PRA applications, sensitivity studies and uncertainty considerations should be incorporated in the PRA evaluations. 4 refs, 3 tabs

HTGR accident and risk assessment

International Nuclear Information System (INIS)

Silady, F.A.; Everline, C.J.; Houghton, W.J.

1982-01-01

This paper is a synopsis of the high-temperature gas-cooled reactor probabilistic risk assessments (PRAs) performed by General Atomic Company. Principal topics presented include: HTGR safety assessments, peer interfaces, safety research, process gas explosions, quantitative safety goals, licensing applications of PRA, enhanced safety, investment risk assessments, and PRA design integration

40 CFR 60.1695 - What happens to the operating requirements during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... requirements during periods of startup, shutdown, and malfunction? 60.1695 Section 60.1695 Protection of... Requirements § 60.1695 What happens to the operating requirements during periods of startup, shutdown, and... municipal waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or...

Attributes of Full Scope Level 1 Probabilistic Safety Assessment (PSA) for Applications in Nuclear Power Plants

International Nuclear Information System (INIS)

2016-10-01

This publication supersedes IAEA-TECDOC-1511, Determining the Quality of Probabilistic Safety Assessment (PSA) for Applications in Nuclear Power Plants (published in 2006), which provided detailed information on technical features of a restricted scope PSA aimed at analysing only internal initiating events caused by random component failures and human errors, and accident sequences that may lead to reactor core damage during operation. The present publication extends the scope of the PSA to cover a broader range of internal and external hazards, and low power and shutdown modes of nuclear power plant operation. In addition, some PSA aspects relevant to lessons learned from the accident at the Fukushima Daiichi nuclear power plant are also considered

Need to use probabilistic risk approach in performance assessment of waste disposal facilities

International Nuclear Information System (INIS)

Bonano, E.J.; Gallegos, D.P.

1991-01-01

Regulations governing the disposal of radioactive, hazardous, and/or mixed wastes will likely require, either directly or indirectly, that the performance of disposal facilities be assessed quantitatively. Such analyses, commonly called ''performance assessments,'' rely on the use of predictive models to arrive at a quantitative estimate of the potential impact of disposal on the environment and the safety and health of the public. It has been recognized that a suite of uncertainties affect the results of a performance assessment. These uncertainties are conventionally categorized as (1) uncertainty in the future state of the disposal system (facility and surrounding medium), (2) uncertainty in models (including conceptual models, mathematical models, and computer codes), and (3) uncertainty in data and parameters. Decisions regarding the suitability of a waste disposal facility must be made in light of these uncertainties. Hence, an approach is needed that would allow the explicit consideration of these uncertainties so that their impact on the estimated consequences of disposal can be evaluated. While most regulations for waste disposal do not prescribe the consideration of uncertainties, it is proposed that, even in such cases, a meaningful decision regarding the suitability of a waste disposal facility cannot be made without considering the impact of the attendant uncertainties. A probabilistic risk assessment (PRA) approach provides the formalism for considering the uncertainties and the technical basis that the decision makers can use in discharging their duties. A PRA methodology developed and demonstrated for the disposal of high-level radioactive waste provides a general framework for assessing the disposal of all types of wastes (radioactive, hazardous, and mixed). 15 refs., 1 fig., 1 tab

Dynamic modeling of the tradeoff between productivity and safety in critical engineering systems

International Nuclear Information System (INIS)

Cowing, Michelle M.; Elisabeth Pate-Cornell, M.; Glynn, Peter W.

2004-01-01

Short-term tradeoffs between productivity and safety often exist in the operation of critical facilities such as nuclear power plants, offshore oil platforms, or simply individual cars. For example, interruption of operations for maintenance on demand can decrease short-term productivity but may be needed to ensure safety. Operations are interrupted for several reasons: scheduled maintenance, maintenance on demand, response to warnings, subsystem failure, or a catastrophic accident. The choice of operational procedures (e.g. timing and extent of scheduled maintenance) generally affects the probabilities of both production interruptions and catastrophic failures. In this paper, we present and illustrate a dynamic probabilistic model designed to describe the long-term evolution of such a system through the different phases of operation, shutdown, and possibly accident. The model's parameters represent explicitly the effects of different components' performance on the system's safety and reliability through an engineering probabilistic risk assessment (PRA). In addition to PRA, a Markov model is used to track the evolution of the system and its components through different performance phases. The model parameters are then linked to different operations strategies, to allow computation of the effects of each management strategy on the system's long-term productivity and safety. Decision analysis is then used to support the management of the short-term trade-offs between productivity and safety in order to maximize long-term performance. The value function is that of plant managers, within the constraints set by local utility commissions and national (e.g. energy) agencies. This model is illustrated by the case of outages (planned and unplanned) in nuclear power plants to show how it can be used to guide policy decisions regarding outage frequency and plant lifetime, and more specifically, the choice of a reactor tripping policy as a function of the state of the

77 FR 73968 - Reconsideration of Certain New Source and Startup/Shutdown Issues: National Emission Standards...

Science.gov (United States)

2012-12-12

...; FRL-9762-1] RIN 2060-AR62 Reconsideration of Certain New Source and Startup/Shutdown Issues: National... Source and Startup/Shutdown Issues: National Emission Standards for Hazardous Air Pollutants from Coal... November 30, 2012, proposed ``Reconsideration of Certain New Source and Startup/Shutdown Issues: National...

Development status of PIUS/ISER - a inherently safe reactor for the international use

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki

1987-01-01

It is just in early 1980s that LWR-based nuclear power has become a substantial power source. Though the safety level of nuclear power is always claimed to be sufficiently high by the industry, it rests on the idea of defense in depth, the calculation by probabilistic risk assessment (PRA) or probabilistic safety assessment (PSA). The TMI-2 and Chernobyl-4 accidents occurred in the industrially most advanced countries. In this paper, an alternative way to safe nuclear power is sought in so-called inherently safe reactors (ISR) including the LWR type PIUS/ISER. With proper consideration into the design of nuclear reactor plants, those can be made basically safe through the use of passive safe mechanism for their design. In short, an ISR is a nuclear power reactor which has passive and intrinsic core cooling capability and automatic shutdown capability. As the nuclear power reactors which are currently claimed to be inherently safe, there are the process inherent and ultimately safe reactor (PIUS) of ASEA-ATOM Sweden and the inherently safe and economical reactor (ISER) of the University of Tokyo, Japan, of LWR type. The current status of the development, the reliability, and some technical problems of ISER/PIUS and the attitude of various countries toward ISER/PIUS are described. (Kako, I.)

40 CFR 60.1220 - What happens to the emission limits during periods of startup, shutdown, and malfunction?

Science.gov (United States)

2010-07-01

... during periods of startup, shutdown, and malfunction? 60.1220 Section 60.1220 Protection of Environment... Emission Limits § 60.1220 What happens to the emission limits during periods of startup, shutdown, and... waste combustion unit startup, shutdown, or malfunction. (b) Each startup, shutdown, or malfunction must...

BWR shutdown analyzer using artificial intelligence (AI) techniques

International Nuclear Information System (INIS)

Cain, D.G.

1986-01-01

A prototype alarm system for detecting abnormal reactor shutdowns based on artificial intelligence technology is described. The system incorporates knowledge about Boiling Water Reactor (BWR) plant design and component behavior, as well as knowledge required to distinguish normal, abnormal, and ATWS accident conditions. The system was developed using a software tool environment for creating knowledge-based applications on a LISP machine. To facilitate prototype implementation and evaluation, a casual simulation of BWR shutdown sequences was developed and interfaced with the alarm system. An intelligent graphics interface for execution and control is described. System performance considerations and general observations relating to artificial intelligence application to nuclear power plant problems are provided

Probabilities of inherent shutdown of unprotected events in innovative liquid metal reactors

International Nuclear Information System (INIS)

Mueller, C.J.; Wade, D.C.

1988-01-01

The uncertainty in predicting the effectiveness of inherent shutdown in innovative liquid metal cooled reactors with metallic fuel results from three broad contributing areas of uncertainty: (1) the inability to exactly predict the frequency of ATWS events with potential to challenge the safety systems and require inherent shutdown; (2) the approximation of representing all such events by a selected set of ''generic scenarios''; and (3) the inability to exactly calculate the core response to the selected generic scenarios. This paper discusses the work being done to address each of these contributing areas, identifies the design and research approaches being used at Argonne National Laboratory to reducing the key contributions to uncertainties in inherent shutdown, and presents results. The conditional probabilities (given ATWS initiation) of achieving temperatures capable of defeating inherent shutdown are shown to range from /approximately/0.1% to negligible for current designs

A probabilistic analysis of rapid boron dilution scenarios

International Nuclear Information System (INIS)

Kohut, P.; Diamond, D.J.

1993-01-01

A probabilistic and deterministic analysis of a rapid boron dilution scenario related to reactor restart was performed. The event is initiated by a loss of off-site power during the startup dilution process. The automatic restart of the charging pump in such cases may lead to the accumulation of a diluted slug of water in the lower plenum. The restart of the reactor coolant pumps may send the diluted slug through the core, adding sufficient reactivity to overcome the shutdown margin and cause a power excursion. The concern is that the power excursion is sufficient in certain circumstances to cause fuel damage. The estimated core damage frequency based on the scoping analysis is 1.0--3.0E-05/yr for the plants analyzed. These are relatively significant values when compared to desirable goals. The analysis contained assumptions related to plant specific design characteristics which may lead to non-conservative estimates. The most important conservative assumptions were that mixing of the injected diluted water is insignificant and that fuel damage occurs when the slug passes through the core

The Bulgaria before shut-down of next two blocks

International Nuclear Information System (INIS)

Dobak, D.

2005-01-01

The Ministry of Trade and Industry of United Kingdom in the frame of realization of programmes for the Middle and East Europe in the area of nuclear energetics during October 5 - 7, 2005 in Kozloduj has organized the Second International Conference on the theme 'Liquidation, social and economic changes'. In this paper author informs about Kozloduj NPP and plans for shut-down of this NPP as well as consequences of the shut-down. One of them the increase of unemployment and social impact for this region are presented

Primary shutdown system monitoring unit for nuclear power plants

International Nuclear Information System (INIS)

Khan, Tahir Kamal; Balasubramanian, R.; Agilandaeswari, K.

2013-01-01

Shut off rods made up of neutron absorbing material are used as Primary Shutdown System. To reduce the power of the reactor under certain abnormal operating conditions, these rods must go down into the core within a specified time. Any malfunctioning in the movement of rods cannot be tolerated and Secondary Shutdown System (SSS) must be actuated within stipulated time to reduce the reactor power. A special safety critical, hardwired electronics unit has been designed to detect failure of PSS Shut off rods movements and generate trip signals for initiating SSS. (author)

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Analysis of core damage frequency from internal events for plant operational state 5 during a refueling outage. Internal events appendices K to M

International Nuclear Information System (INIS)

Forester, J.; Yakle, J.; Walsh, B.; Darby, J.; Whitehead, D.; Staple, B.; Brown, T.

1994-07-01

This report provides supporting documentation for various tasks associated with the performance of the probabilistic risk assessment for Plant Operational State 5 (approximately Cold Shutdown as defined by Grand Gulf Technical Specifications) during a refueling outage at Grand Gulf, Unit 1 as documented in Volume 2, Part 1 of NUREG/CR-6143. The report contains the following appendices: K - HEP Locator Files; L - Supporting Information for the Plant Damage State Analysis; M - Summary of Results from the Coarse Screening Analysis - Phase 1A

Multi-unit shutdown due to boiler feedwater chemical excursion

International Nuclear Information System (INIS)

Diebel, M.E.

1991-01-01

Ontario Hydro's Bruce Nuclear Generating Station 'B' consists of four 935 W CANDU units located on the east shore of Lake Huron in the province of Ontario, Canada. On July 25 and 26, 1989 three of the four operating units were shutdown due to boiler feedwater chemical excursions initiated by a process upset in the Water Treatment Plant that provides demineralized make-up water to all four units. The chemicals that escaped from an ion exchange vessel during a routine regeneration very quickly spread through the condensate make-up system and into the boiler feedwater systems. This resulted in boiler sulfate levels exceeding shutdown limits. A total of 260 GWH of electrical generation was unexpectedly made unavailable to the grid at a time of peak seasonal demand. This event exposed several unforeseen deficiencies and vulnerabilities in the automatic demineralized water make-up quality protection scheme, system designs, operating procedures and the ability of operating personnel to recognize and appropriately respond to such an event. The combination of these factors contributed towards turning a minor system upset into a major multi-unit shutdown. This paper provides the details of the actual event initiation in the Water Treatment Plant and describes the sequence of events that led to the eventual shutdown of three units and near shutdown of the fourth. The design inadequacies, procedural deficiencies and operating personnel responses and difficulties are described. The process of recovering from this event, the flushing out of system piping, boilers and the feedwater train is covered as well as our experiences with setting up supplemental demineralized water supplies including trucking in water and the use of rental trailer mounted demineralizing systems. System design, procedural and operational changes that have been made and that are still being worked on in response to this event are described. The latest evidence of the effect of this event on boiler tube

Standardization of the time for the execution of HANARO start-up and shutdown procedures

International Nuclear Information System (INIS)

Choi, H. Y.; Lim, I. C.; Hwang, S. R.; Kang, T. J.; Youn, D. B.

2003-01-01

For the standardization of the time to execute HANARO start-up and shutdown procedures, code names were assigned to the individual procedures and the work time were investigated. The data recorded by the operators during start-up and shutdown were statistically analyzed. The analysis results will be used for the standardization of start-up and shutdown procedures and it will be reflected in the procedure document

Optimal test intervals for shutdown systems for the Cernavoda nuclear power station

International Nuclear Information System (INIS)

Negut, Gh.; Laslau, F.

1993-01-01

Cernavoda nuclear power station required a complete PSA study. As a part of this study, an important goal to enhance the effectiveness of the plant operation is to establish optimal test intervals for the important engineering safety systems. The paper presents, briefly, the current methods to optimize the test intervals. For this reason it was used Vesely methods to establish optimal test intervals and Frantic code to survey the influence of the test intervals on system availability. The applications were done on the Shutdown System no. 1, a shutdown system provided whit solid rods and on Shutdown System no. 2 provided with injecting poison. The shutdown systems receive nine total independent scram signals that dictate the test interval. Fault trees for the both safety systems were developed. For the fault tree solutions an original code developed in our Institute was used. The results, intended to be implemented in the technical specifications for test and operation of Cernavoda NPS are presented