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 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

Probabilistic risk assessment (PRA): status report and guidance for regulatory application. Draft report for comment

International Nuclear Information System (INIS)

1984-02-01

This document describes the current status of the methodologies used in probabilistic risk assessment (PRA) and provides guidance for the application of the results of PRAs to the nuclear reactor regulatory process. The PRA studies that have been completed or are underway are reviewed. The levels of maturity of the methodologies used in a PRA are discussed. Insights derived from PRAs are listed. The potential uses of PRA results for regulatory purposes are discussed

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

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

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)

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.

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)

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

Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Technical Exchange Meeting

Energy Technology Data Exchange (ETDEWEB)

Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2013-09-01

During FY13, the INL developed an advanced SMR PRA framework which has been described in the report Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Detailed Technical Framework Specification, INL/EXT-13-28974 (April 2013). In this framework, the various areas are considered: Probabilistic models to provide information specific to advanced SMRs Representation of specific SMR design issues such as having co-located modules and passive safety features Use of modern open-source and readily available analysis methods Internal and external events resulting in impacts to safety All-hazards considerations Methods to support the identification of design vulnerabilities Mechanistic and probabilistic data needs to support modeling and tools In order to describe this framework more fully and obtain feedback on the proposed approaches, the INL hosted a technical exchange meeting during August 2013. This report describes the outcomes of that meeting.

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

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

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

Safety analysis, risk assessment, and risk acceptance criteria

International Nuclear Information System (INIS)

Jamali, K.

1997-01-01

This paper discusses a number of topics that relate safety analysis as documented in the Department of Energy (DOE) safety analysis reports (SARs), probabilistic risk assessments (PRA) as characterized primarily in the context of the techniques that have assumed some level of formality in commercial nuclear power plant applications, and risk acceptance criteria as an outgrowth of PRA applications. DOE SARs of interest are those that are prepared for DOE facilities under DOE Order 5480.23 and the implementing guidance in DOE STD-3009-94. It must be noted that the primary area of application for DOE STD-3009 is existing DOE facilities and that certain modifications of the STD-3009 approach are necessary in SARs for new facilities. Moreover, it is the hazard analysis (HA) and accident analysis (AA) portions of these SARs that are relevant to the present discussions. Although PRAs can be qualitative in nature, PRA as used in this paper refers more generally to all quantitative risk assessments and their underlying methods. HA as used in this paper refers more generally to all qualitative risk assessments and their underlying methods that have been in use in hazardous facilities other than nuclear power plants. This discussion includes both quantitative and qualitative risk assessment methods. PRA has been used, improved, developed, and refined since the Reactor Safety Study (WASH-1400) was published in 1975 by the Nuclear Regulatory Commission (NRC). Much debate has ensued since WASH-1400 on exactly what the role of PRA should be in plant design, reactor licensing, 'ensuring' plant and process safety, and a large number of other decisions that must be made for potentially hazardous activities. Of particular interest in this area is whether the risks quantified using PRA should be compared with numerical risk acceptance criteria (RACs) to determine whether a facility is 'safe.' Use of RACs requires quantitative estimates of consequence frequency and magnitude

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

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

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)

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.

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)

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

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

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

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.

76 FR 81998 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2011-12-29

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY..., ``Methodology for Low Power/Shutdown Fire PRA--Draft Report for Comment.'' DATES: Submit comments by March 01... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

Probabilistic risk assessment (PRA) update in light of the accident at Fukushima Daiichi Nuclear Power Station - 15461

International Nuclear Information System (INIS)

Maeda, K.; Abe, H.; Hirokawa, N.; Satou, C.

2015-01-01

We have performed internal and external event probabilistic risk assessments (PRA) for boiling water reactor power nuclear plants to identify the important accident sequence groups and to evaluate the effectiveness of the additional severe accident measures, regarding to the new regulatory requirements implemented after the accident at Fukushima Daiichi Nuclear Power Station in Japan in 2011. In addition, we will further update our PRA by extracting problems and improvements from the current PRA, by catching up the state-of-the-art knowledge, modern PRA methodologies in order to contribute voluntarily to safety improvement as well as to comply with regulations. In this document, prior to the extensive PRA updates, we would describe technical contents and qualitative results about PRA updates that have been performed preliminary so far, especially about the external event (seismic) PRA and how to model the additionally deployed severe accident measures (e.g. power supply car, fire engine) so that they can be function external hazards, such as component failure rate of equipment, human reliability 'out of control room', and mission time extension. (authors)

77 FR 10576 - Methodology for Low Power/Shutdown Fire PRA

Science.gov (United States)

2012-02-22

... NUCLEAR REGULATORY COMMISSION [NRC-2011-0295] Methodology for Low Power/Shutdown Fire PRA AGENCY.../Shutdown Fire PRA.'' In response to request from members of the public, the NRC is extending the public... risk assessment (PRA) method for quantitatively analyzing fire risk in commercial nuclear power plants...

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

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.

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

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

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

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

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

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)

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

Evaluation of allowed outage time using PRA results

International Nuclear Information System (INIS)

Johanson, G.

1985-01-01

In a probabilistic risk assessment (PRA) different measures of risk importance can be established. These measures can be used as a basis for further evaluation and determination of allowed outage time for specific components, within safety systems of a nuclear power plant. In order to optimize the allowed outage time (AOT) stipulated in the plant's Technical Specification it is necessary to create a methodology which could incorporate existing PRA data into a quantitative extrapolation. In order to evaluate the plant risk status due to AOT in a quantitative manner, the risk achievement worth is utilized. Risk achievement worth is defined as follows: to measure the worth of a feature, in achieving the present risk, one approach is to remove the feature and then determine how much the risk has increased. Thus, the risk achievement worth is formally defined to be the increase in risk if the feature were assumed not be there or to be failed. Another parameter of interest for this analysis is the shutdown risk increase. The shutdown risk achievement worth must be incorporated into the accident sequence risk achievement worth to arrive at an optimal set of plant specific AOTs

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.

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

Economic impact assessment in pest risk analysis

NARCIS (Netherlands)

Soliman, T.A.A.; Mourits, M.C.M.; Oude Lansink, A.G.J.M.; Werf, van der W.

2010-01-01

According to international treaties, phytosanitary measures against introduction and spread of invasive plant pests must be justified by a science-based pest risk analysis (PRA). Part of the PRA consists of an assessment of potential economic consequences. This paper evaluates the main available

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

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

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

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

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

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

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

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.)

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

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)

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.

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

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)

A PRA case study of extended long term decay heat removal for shutdown risk assessment

International Nuclear Information System (INIS)

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

1992-01-01

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor II (EBR-II), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL). The results of this PRA have shown that the decay heat removal system for EBR-II is extremely robust and reliable. In addition, the methodology used demonstrates how the actions of other systems not normally used for actions of other systems not normally used for decay heat removal can be used to expand the mission time of the decay heat removal system and further increase its reliability. The methodology may also be extended to account for the impact of non-safety systems in enhancing the reliability of other dedicated safety systems

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)

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

Value impact analysis utilizing PRA techniques combined with a hybrid plant model

International Nuclear Information System (INIS)

Edson, J.L.; Stillwell, D.W.

1989-01-01

A value impact analysis (VIA) has been performed by the INEL to support a NRC Regulatory Analysis for resolution of Generic Issue (GI) 29, Bolting Degradation or Failure in Nuclear Power Plants. A VIA for replacing the reactor coolant pressure boundary (RCPB) bolts of BWRs and PWRs was previously prepared by Pacific Northwest Laboratories in 1985 under instructions limiting the VIA to the potential for failure of primary pressure boundary bolting. Subsequently the INEL was requested to perform a VIA that included non primary systems and component support bolts to be compatible with the resolution of the broader issue. Because the initial list of systems and bolting applications that could be included in the VIA was very large, including them all in the VIA would likely result in analyzing some that have little if any effect on public risk. This paper discusses how PRA techniques combined with a hybrid plant model were used to determine which bolts have the potential to be significant contributors to public risk if they were to fail, and therefore were included in the VIA

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.

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

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

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

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

Cable Hot Shorts and Circuit Analysis in Fire Risk Assessment

International Nuclear Information System (INIS)

LaChance, Jeffrey; Nowlen, Steven P.; Wyant, Frank

1999-01-01

Under existing methods of probabilistic risk assessment (PRA), the analysis of fire-induced circuit faults has typically been conducted on a simplistic basis. In particular, those hot-short methodologies that have been applied remain controversial in regards to the scope of the assessments, the underlying methods, and the assumptions employed. To address weaknesses in fire PRA methodologies, the USNRC has initiated a fire risk analysis research program that includes a task for improving the tools for performing circuit analysis. The objective of this task is to obtain a better understanding of the mechanisms linking fire-induced cable damage to potentially risk-significant failure modes of power, control, and instrumentation cables. This paper discusses the current status of the circuit analysis task

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)

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

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

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)

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)

System Analysis and Risk Assessment (SARA) system

International Nuclear Information System (INIS)

Krantz, E.A.; Russell, K.D.; Stewart, H.D.; Van Siclen, V.S.

1986-01-01

Utilization of Probabilistic Risk Assessment (PRA) related information in the day-to-day operation of plant systems has, in the past, been impracticable due to the size of the computers needed to run PRA codes. This paper discusses a microcomputer-based database system which can greatly enhance the capability of operators or regulators to incorporate PRA methodologies into their routine decision making. This system is called the System Analysis and Risk Assessment (SARA) system. SARA was developed by EG and G Idaho, Inc. at the Idaho National Engineering Laboratory to facilitate the study of frequency and consequence analyses of accident sequences from a large number of light water reactors (LWRs) in this country. This information is being amassed by several studies sponsored by the United States Nuclear Regulatory Commission (USNRC). To meet the need of portability and accessibility, and to perform the variety of calculations necessary, it was felt that a microcomputer-based system would be most suitable

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

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

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)

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)

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

Results and insights of a level-1 internal event PRA of a PWR during mid-loop operations

International Nuclear Information System (INIS)

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

1993-01-01

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 analysis 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. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed

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)

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.

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

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

Full text of publication follows: The pressurized light water cooled, medium power (1000 MWt) IRIS (International Reactor Innovative and Secure) has been under development for four years by an international consortium of over 21 organizations from ten countries. The plant conceptual design was completed in 2001 and the preliminary design is nearing completion. The pre-application licensing process with NRC started in October, 2002. IRIS has been primarily focused on establishing a design with innovative safety characteristics. The first line of defense in IRIS is to eliminate event initiators that could potentially lead to core damage. In IRIS, this concept is implemented through the 'safety by design' approach, which allows to minimize the number and complexity of the safety systems and required operator actions. The end result is a design with significantly reduced complexity and improved operability, and extensive plant simplifications to enhance construction. To support the optimization of the plant design and confirm the effectiveness of the safety by design approach in mitigating or eliminating events and thus providing a significant reduction in the probability of severe accidents, the PRA is being used as an integral part of the design process. A preliminary but extensive Level 1 PRA model has been developed to support the pre-application licensing of the IRIS design. As a result of the Preliminary IRIS PRA, an optimization of the design from a reliability point of view was completed, and an extremely low (about 1.2 E -8 ) core damage frequency (CDF) was assessed to confirm the impact of the safety by design approach. This first assessment is a result of a PRA model including internal initiating events. During this assessment, several assumptions were necessary to complete the CDF evaluation. In particular Anticipated Transients Without Scram (ATWS) were not included in this initial assessment, because their contribution to core damage frequency was assumed

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

Analysis of dependent failures in risk assessment and reliability evaluation

International Nuclear Information System (INIS)

Fleming, K.N.; Mosleh, A.; Kelley, A.P. Jr.; Gas-Cooled Reactors Associates, La Jolla, CA)

1983-01-01

The ability to estimate the risk of potential reactor accidents is largely determined by the ability to analyze statistically dependent multiple failures. The importance of dependent failures has been indicated in recent probabilistic risk assessment (PRA) studies as well as in reports of reactor operating experiences. This article highlights the importance of several different types of dependent failures from the perspective of the risk and reliability analyst and provides references to the methods and data available for their analysis. In addition to describing the current state of the art, some recent advances, pitfalls, misconceptions, and limitations of some approaches to dependent failure analysis are addressed. A summary is included of the discourse on this subject, which is presented in the Institute of Electrical and Electronics Engineers/American Nuclear Society PRA Procedures Guide

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 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

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.

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)

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

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

Top event prevention analysis: A deterministic use of PRA

International Nuclear Information System (INIS)

Worrell, R.B.; Blanchard, D.P.

1996-01-01

This paper describes the application of Top Event Prevention Analysis. The analysis finds prevention sets which are combinations of basic events that can prevent the occurrence of a fault tree top event such as core damage. The problem analyzed in this application is that of choosing a subset of Motor-Operated Valves (MOVs) for testing under the Generic Letter 89-10 program such that the desired level of safety is achieved while providing economic relief from the burden of testing all safety-related valves. A brief summary of the method is given, and the process used to produce a core damage expression from Level 1 PRA models for a PWR is described. The analysis provides an alternative to the use of importance measures for finding the important combination of events in a core damage expression. This application of Top Event Prevention Analysis to the MOV problem was achieve with currently available software

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

Dependent failure analysis research for the US NRC Risk Methods Integration and Evaluation Program

International Nuclear Information System (INIS)

Bohn, M.P.; Stack, D.W.; Campbell, D.J.; Rooney, J.J.; Rasmuson, D.M.

1985-01-01

The Risk Methods Integration and Evaluation Program (RMIEP), which is being performed for the Nuclear Regulatory Commission by Sandia National Laboratories, has the goals of developing new risk assessment methods and integrating the new and existing methods in a uniform procedure for performing an in-depth probabilistic risk assessment (PRA) with consistent levels of analysis for internal, external, and dependent failure scenarios. An important part of RMIEP is the recognition of the crucial importance of dependent common cause failures (CCFs) and the pressing need to develop effective methods for analyzing CCFs as part of a PRA. The NRC-sponsored Integrated Dependent Failure Methodology Program at Sandia is addressing this need. This paper presents a preliminary approach for analyzing CCFs as part of a PRA. A nine-step procedure for efficiently screening and analyzing dependent failure scenarios is presented, and each step is discussed

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)

Results and insights of a level-1 internal event PRA of a PWR during mid-loop operations

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.; Holmes, B.; Su, R.F.; Dang, V.; Siu, N.; Bley, D.; Johnson, D.; Lin, J.

1994-01-01

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 analysis 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 BNL and SNL. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed

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

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

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 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

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)

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)

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)

Modelling and mapping spread in pest risk analysis: a generic approach

NARCIS (Netherlands)

Kehlenbeck, H.; Robinet, C.; Werf, van der W.; Kriticos, D.; Reynaud, P.; Baker, R.

2012-01-01

Assessing the likelihood and magnitude of spread is one of the cornerstones of pest risk analysis (PRA), and is usually based on qualitative expert judgment. This paper proposes a suite of simple ecological models to support risk assessors who also wish to estimate the rate and extent of spread,

Nuclear power plant risk assembly and decomposition for risk management

International Nuclear Information System (INIS)

Iden, D.C.

1985-01-01

The state-of-the-art method for analyzing the risk from nuclear power plants is probabilistic risk assessment (PRA). The intermediate results of a PRA are first assembled to quantify the risk from operating a nuclear power plant in the form of (1) core damage (or core melt) frequency, (2) plant damage state frequencies, (3) release category frequencies, and (4) the frequency of exceeding specific levels of offsite consequences. Once the overall PRA results have been quantified, the next step is to decompose those results into the individual contributors to each of the four forms of risk in some rank order. The way in which the PRA model is set up to assemble and decompose the plant risk determines the ease and usefulness of the PRA model as a risk management tool for evaluating perturbations to the PRA model. These perturbations can take the form of technical specification changes, hardware modifications, procedural changes, etc. The matrix formalism developed by Dr. Stan Kaplan for risk assembly and decomposition represents a significant breakthrough in making the PRA model an effective risk management tool. The key to understanding the matrix formalism and making it a useful tool for managing nuclear power plant risk is the structure of the PRA model. PRA risk model structure and decomposition of the risk results are discussed with the Seabrook PRA as an example

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

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)

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

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)

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

Formalized search strategies for human risk contributions

International Nuclear Information System (INIS)

Rasmussen, J.; Pedersen, O.M.

1982-07-01

For risk management, the results of a probabilistic risk analysis (PRA) as well as the underlying assumptions can be used as references in a closed-loop risk control; and the analyses of operational experiences as a means of feedback. In this context, the need for explicit definition and documentation of the PRA coverage, including the search strategies applied, is discussed and aids are proposed such as plant description in terms of a formal abstraction hierarchy and use of cause-consequence-charts for the documentation of not only the results of PRA but also of its coverage. Typical human risk contributions are described on the basis of general plant design features relevant for risk and accident analysis. With this background, search strategies for human risk contributions are treated: Under the designation ''work analysis'', procedures for the analysis of familiar, well trained, planned tasks are proposed. Strategies for identifying human risk contributions outside this category are outlined. (author)

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

Workshop on the use of PRA methodology for the analysis of reactor events and operational data: Proceedings

International Nuclear Information System (INIS)

Rasmuson, D.M.

1992-06-01

A workshop entitled ''The Use of PRA Methodology for the Analysis of Reactor Events and Operational Data'' was held on January 29--30, 1992 in Annapolis, Maryland. Over 50 participants from the NRC, its contractors, and others participated in the meetings. During the first day, presentations were made by invited speakers to discuss issues in relevant topics. On the second day, discussion groups were held to focus on three areas: risk significance of operational events, industry risk profile and generic concerns, and risk monitoring and risk-based performance indicators. Important considerations identified from the workshop are the following: Improve the Accident Sequence Precursor models and data. Improve the SCSS and NPRDS (e.g., by adding detailed performance information on selected components, by improving narratives on failure causes). Develop risk-based performance indicators. Use risk insights to help focus trending and performance analyses of components, systems, initiators, and sequences. Improve the statistical quality of trending and performance analyses. Flag implications of special conditions (e.g., external events, containment performance) during data studies. Trend common cause and human performance using appropriate models to obtain a better understanding of the impact and causes of failure. Develop a method for producing an industry risk profile

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

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)

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

Human Reliability Analysis in Support of Risk Assessment for Positive Train Control

Science.gov (United States)

2003-06-01

This report describes an approach to evaluating the reliability of human actions that are modeled in a probabilistic risk assessment : (PRA) of train control operations. This approach to human reliability analysis (HRA) has been applied in the case o...

Risk-Informed External Hazards Analysis for Seismic and Flooding Phenomena for a Generic PWR

Energy Technology Data Exchange (ETDEWEB)

Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Prescott, Steve [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ma, Zhegang [Idaho National Lab. (INL), Idaho Falls, ID (United States); Spears, Bob [Idaho National Lab. (INL), Idaho Falls, ID (United States); Szilard, Ronaldo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Coleman, Justin [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kosbab, Ben [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-07-26

This report describes the activities performed during the FY2017 for the US-DOE Light Water Reactor Sustainability Risk-Informed Safety Margin Characterization (LWRS-RISMC), Industry Application #2. The scope of Industry Application #2 is to deliver a risk-informed external hazards safety analysis for a representative nuclear power plant. Following the advancements occurred during the previous FYs (toolkits identification, models development), FY2017 focused on: increasing the level of realism of the analysis; improving the tools and the coupling methodologies. In particular the following objectives were achieved: calculation of buildings pounding and their effects on components seismic fragility; development of a SAPHIRE code PRA models for 3-loops Westinghouse PWR; set-up of a methodology for performing static-dynamic PRA coupling between SAPHIRE and EMRALD codes; coupling RELAP5-3D/RAVEN for performing Best-Estimate Plus Uncertainty analysis and automatic limit surface search; and execute sample calculations for demonstrating the capabilities of the toolkit in performing a risk-informed external hazards safety analyses.

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

Technical requirements for the ASME PRA standard for nuclear power plant applications

International Nuclear Information System (INIS)

Fleming, Karl N.; Bernsen, Sidney A.; Simard, Ronald L.

2000-01-01

In 1998 the American Society of Mechanical Engineers (ASME) formed the Committee on Nuclear Risk Management (CNRM) and a Project Team to develop a standard on PRAs for use in risk informed applications. This ASME standard is being developed to help provide an adequate level of quality in PRAs that are being used to support ASME initiatives to risk informed in-service inspection (ISI) and in-service testing (IST) of nuclear power plant components. A related need supported by the industry and the U.S. Nuclear Regulatory Commission is to reduce the level of effort that is being expended in pilot applications of risk informed initiatives to address questions about the sufficiency of quality in the supporting PRA models. The purpose of this paper is to discuss the authors' views on some of the technical issues that were encountered in the effort to develop the ASME PRA standard. Draft 12 of this standard has been issued for comment, and is currently being finalized with the aim of releasing the standard in early 2001. (author)

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.)

Johnson Space Center's Risk and Reliability Analysis Group 2008 Annual Report

Science.gov (United States)

Valentine, Mark; Boyer, Roger; Cross, Bob; Hamlin, Teri; Roelant, Henk; Stewart, Mike; Bigler, Mark; Winter, Scott; Reistle, Bruce; Heydorn,Dick

2009-01-01

The Johnson Space Center (JSC) Safety & Mission Assurance (S&MA) Directorate s Risk and Reliability Analysis Group provides both mathematical and engineering analysis expertise in the areas of Probabilistic Risk Assessment (PRA), Reliability and Maintainability (R&M) analysis, and data collection and analysis. The fundamental goal of this group is to provide National Aeronautics and Space Administration (NASA) decisionmakers with the necessary information to make informed decisions when evaluating personnel, flight hardware, and public safety concerns associated with current operating systems as well as with any future systems. The Analysis Group includes a staff of statistical and reliability experts with valuable backgrounds in the statistical, reliability, and engineering fields. This group includes JSC S&MA Analysis Branch personnel as well as S&MA support services contractors, such as Science Applications International Corporation (SAIC) and SoHaR. The Analysis Group s experience base includes nuclear power (both commercial and navy), manufacturing, Department of Defense, chemical, and shipping industries, as well as significant aerospace experience specifically in the Shuttle, International Space Station (ISS), and Constellation Programs. The Analysis Group partners with project and program offices, other NASA centers, NASA contractors, and universities to provide additional resources or information to the group when performing various analysis tasks. The JSC S&MA Analysis Group is recognized as a leader in risk and reliability analysis within the NASA community. Therefore, the Analysis Group is in high demand to help the Space Shuttle Program (SSP) continue to fly safely, assist in designing the next generation spacecraft for the Constellation Program (CxP), and promote advanced analytical techniques. The Analysis Section s tasks include teaching classes and instituting personnel qualification processes to enhance the professional abilities of our analysts

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.

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

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)

A Regional Decision Support Scheme for Pest Risk Analysis in Southeast Asia.

Science.gov (United States)

Soliman, T; MacLeod, A; Mumford, J D; Nghiem, T P L; Tan, H T W; Papworth, S K; Corlett, R T; Carrasco, L R

2016-05-01

A key justification to support plant health regulations is the ability of quarantine services to conduct pest risk analyses (PRA). Despite the supranational nature of biological invasions and the close proximity and connectivity of Southeast Asian countries, PRAs are conducted at the national level. Furthermore, some countries have limited experience in the development of PRAs, which may result in inadequate phytosanitary responses that put their plant resources at risk to pests vectored via international trade. We review existing decision support schemes for PRAs and, following international standards for phytosanitary measures, propose new methods that adapt existing practices to suit the unique characteristics of Southeast Asia. Using a formal written expert elicitation survey, a panel of regional scientific experts was asked to identify and rate unique traits of Southeast Asia with respect to PRA. Subsequently, an expert elicitation workshop with plant protection officials was conducted to verify the potential applicability of the developed methods. Rich biodiversity, shortage of trained personnel, social vulnerability, tropical climate, agriculture-dependent economies, high rates of land-use change, and difficulties in implementing risk management options were identified as challenging Southeast Asian traits. The developed methods emphasize local Southeast Asian conditions and could help support authorities responsible for carrying out PRAs within the region. These methods could also facilitate the creation of other PRA schemes in low- and middle-income tropical countries. © 2016 Society for Risk Analysis.

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.

Risk analysis of environmental hazards at the High Flux Beam Reactor

International Nuclear Information System (INIS)

Boccio, J.L.; Ho, V.S.; Johnson, D.H.

1994-01-01

In the late 1980s, a Level 1 internal event probabilistic risk assessment (PRA) was performed for the High-Flux Beam Reactor (HFBR), a US Department of Energy research reactor located at Brookhaven National Laboratory. Prior to the completion of that study, a level 1 PRA for external events was initiated, including environmental hazards such as fire, internal flooding, etc. Although this paper provides a brief summary of the risks from environmental hazards, emphasis will be placed on the methodology employed in utilizing industrial event databases for event frequency determination for the HFBR complex. Since the equipment in the HFBR is different from that of, say, a commercial nuclear power plant, the current approach is to categorize the industrial events according to the hazard initiators instead of categorizing by initiator location. But first a general overview of the analysis

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)

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

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

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

System Analysis and Risk Assessment system (SARA) Version 4.0

International Nuclear Information System (INIS)

Sattison, M.B.; Russell, K.D.; Skinner, N.L.

1992-01-01

This NUREG is the tutorial for the System Analysis and Risk Assessment System (SARA) Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons are provided that walk the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis

Probabilistic risk assessment course documentation. Volume 2. Probability and statistics for PRA applications

International Nuclear Information System (INIS)

Iman, R.L.; Prairie, R.R.; Cramond, W.R.

1985-08-01

This course is intended to provide the necessary probabilistic and statistical skills to perform a PRA. Fundamental background information is reviewed, but the principal purpose is to address specific techniques used in PRAs and to illustrate them with applications. Specific examples and problems are presented for most of the topics

Development of infrastructure for the regulatory authority to implement risk-informed regulation

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

It is important to assure the technical adequacy of probabilistic risk assessment (PRA) to implement risk-informed regulation of nuclear power plants (NPPs). JNES has been conducting various activities, such as development of PRA model, method, and data base, in order to assure the technical adequacy of PRA as development of the infrastructure for the regulatory authority to implement risk-informed regulation. In 2012, JNES updated the reliability data base used in PRA and improved PRA models to enhance the technical bases of PRA. In addition, JNES has been establishing the PRA model for fuel damage in the spent fuel storage pool in NPPs. As for improvement of PRA model for core damage in reactor, JNES conducted the study including feasibility of a simplified reliability model for digital I and C system developed by the digital I and C task group of OECD/NEA CSNI WGRISK by reproducing the sample calculation, and improvement of PRA models of individual NPPs in Japan. JNES is making effort to develop the procedures of internal fire PRA and internal flooding PRA. To improve the internal fire PRA, JNES is participating in OECD/NEA FIRE project to obtain the latest information and to validate and improve the fire propagation analysis codes and the parameters. JNES is establishing a method for analyzing internal influence due to flooding in NPPs, and this method is the base to develop the procedure of internal flooding PRA. (author)

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.

System Analysis and Risk Assessment System (SARA), Version 4.0

International Nuclear Information System (INIS)

Russell, K.D.; Sattison, M.B.; Skinner, N.L.; Stewart, H.D.; Wood, S.T.

1992-02-01

This NUREG is the reference manual for the System Analysis and Risk Assessment (SARA) System Version 4.0, a microcomputer-based system used to analyze the safety issues of a family [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. The SARA data base contains PRA data for the dominant accident sequences of a family and descriptive information about the family including event trees, fault trees, and system model diagrams. The number of facility data bases that can be accessed is limited only by the amount of disk storage available. To simulate changes to family systems, SARA users change the failure rates of initiating and basic events and/or modify the structure of the cut sets that make up the event trees, fault trees, and systems. The user then evaluates the effects of these changes through the recalculation of the resultant accident sequence probabilities and importance measures. The results are displayed in tables and graphs

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)

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

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)

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.

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

Integrated Level 3 risk assessment for the LaSalle Unit 2 nuclear power plant

International Nuclear Information System (INIS)

Payne, A.C. Jr.; Brown, T.D.; Miller, L.A.

1991-01-01

An integrated Level 3 probabilistic risk assessment (PRA) was performed on the LaSalle County Station nuclear power plant using state-of-the-art PRA analysis techniques. The objective of this study was to provide an estimate of the risk to the offsite population during full power operation of the plant and to include a characterization of the uncertainties in the calculated risk values. Uncertainties were included in the accident frequency analysis, accident progression analysis, and the source term analysis. Only weather uncertainties were included in the consequence analysis. In this paper selected results from the accident frequency, accident progression, source term, consequence, and integrated risk analyses are discussed and the methods used to perform a fully integrated Level 3 PRA are examined. LaSalle County Station is a two-unit nuclear power plant located 55 miles southwest of Chicago, Illinois. Each unit utilizes a Mark 2 containment to house a General Electric 3323 MWt BWR-5 reactor. This PRA, which was performed on Unit 2, included internal as well as external events. External events that were propagated through the risk analysis included earthquakes, fires, and floods. The internal event accident scenarios included transients, transient-induced LOCAs (inadvertently stuck open relief valves), anticipated transients without scram, and loss of coolant accidents

Organization of Risk Analysis Codes for Living Evaluations (ORACLE)

International Nuclear Information System (INIS)

Batt, D.L.; MacDonald, P.E.; Sattison, M.B.; Vesely, E.

1987-01-01

ORACLE (Organization of Risk Analysis Codes for Living Evaluations) is an integration concept for using risk-based information in United States Nuclear Regulatory Commission (USNRC) applications. Portions of ORACLE are being developed at the Idaho Nationale Engineering Laboratory for the USNRC. The ORACLE concept consists of related databases, software, user interfaces, processes, and quality control checks allowing a wide variety of regulatory problems and activities to be addressed using current, updated PRA information. The ORACLE concept provides for smooth transitions between one code and the next without pre- or post-processing. (orig.)

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

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

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.

Living PRAs [probabilistic risk analysis] made easier with IRRAS [Integrated Reliability and Risk Analysis System

International Nuclear Information System (INIS)

Russell, K.D.; Sattison, M.B.; Rasmuson, D.M.

1989-01-01

The Integrated Reliability and Risk Analysis System (IRRAS) is an integrated PRA software tool that gives the user the ability to create and analyze fault trees and accident sequences using an IBM-compatible microcomputer. This program provides functions that range from graphical fault tree and event tree construction to cut set generation and quantification. IRRAS contains all the capabilities and functions required to create, modify, reduce, and analyze event tree and fault tree models used in the analysis of complex systems and processes. IRRAS uses advanced graphic and analytical techniques to achieve the greatest possible realization of the potential of the microcomputer. When the needs of the user exceed this potential, IRRAS can call upon the power of the mainframe computer. The role of the Idaho National Engineering Laboratory if the IRRAS program is that of software developer and interface to the user community. Version 1.0 of the IRRAS program was released in February 1987 to prove the concept of performing this kind of analysis on microcomputers. This version contained many of the basic features needed for fault tree analysis and was received very well by the PRA community. Since the release of Version 1.0, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version is designated ''IRRAS 2.0''. Version 3.0 will contain all of the features required for efficient event tree and fault tree construction and analysis. 5 refs., 26 figs

Structural Health Monitoring Analysis for the Orbiter Wing Leading Edge

Science.gov (United States)

Yap, Keng C.

2010-01-01

This viewgraph presentation reviews Structural Health Monitoring Analysis for the Orbiter Wing Leading Edge. The Wing Leading Edge Impact Detection System (WLE IDS) and the Impact Analysis Process are also described to monitor WLE debris threats. The contents include: 1) Risk Management via SHM; 2) Hardware Overview; 3) Instrumentation; 4) Sensor Configuration; 5) Debris Hazard Monitoring; 6) Ascent Response Summary; 7) Response Signal; 8) Distribution of Flight Indications; 9) Probabilistic Risk Analysis (PRA); 10) Model Correlation; 11) Impact Tests; 12) Wing Leading Edge Modeling; 13) Ascent Debris PRA Results; and 14) MM/OD PRA Results.

PRATIQUE: a research project to enhance pest risk analysis techniques in the European Union

NARCIS (Netherlands)

Baker, R.H.A.; Battisti, A.; Bremmer, J.; Kenis, M.; Mumford, J.; Petter, F.; Schrader, G.; Bacher, S.; DeBarro, P.; Hulme, P.E.; Karadjova, O.; Oude Lansink, A.; Pruvost, O.; Pysek, P.; Roques, A.; Baranchikov, Y.; Sun, J.H.

2009-01-01

PRATIQUE is an EC-funded 7th Framework research project designed to address the major challenges for pest risk analysis (PRA) in Europe. It has three principal objectives: (a) to assemble the datasets required to construct PRAs valid for the whole of the EU, (b) to conduct multi-disciplinary

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)

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.

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)

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

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

Development of fire PRA methodologies for the analysis of typical Italian NPP designs

International Nuclear Information System (INIS)

Silvestri, E.; Dore, B.; Ferro, G.; Apostolakis, G.

1987-01-01

To compute fire induced Core Melt probability, the results of hazard and propagation analyses were combined with the Core Melt frequency computed for the initiating event and the support state as determined by the fire considered. From the PRA for internal event, the average value of this frequency was found 2.5x10 -3 event/year. Using the average fire frequency the resulting fire induced Core Melt frequency is 1.4x10 -8 event/year. Although high separation of safety systems is required in Italian PWR plants, the frequency of fire induced Core Melt can reach values not negligible with respect to Italian safety standards. For this reason, fire PRA studies for the entire plant are considered necessary and should be performed with appropriate modifications of the methods used for the American plants in order to be able to estimate lower fire induced Core Melt frequencies. (orig./HP)

Analysis of Risk Optimization on the Industrial Area Around

International Nuclear Information System (INIS)

Sony, DT; Demon-Handoyo

2000-01-01

Indonesia as an industrial country, there are large industrial area whichis directly or indirectly have an effect to human health by routine dischargeof waste from industrial installations. So, the criteria limit must bedetermined to regulate industrial area. The PRA method (Probabilistic RiskAssessment) is used in the nuclear technology especially reactor safetytechnology could be applied to accommodate those problems. The principles ofPRA method is to determine probability and consequences for accident ofindustrial plant or transportation of product. The analysis procedureincludes classification of industry activity type and inventories, estimationof external consequences, estimation of probability for installation andtransportation accident, determination of social risk and prioritization ofrisks. Calculation of consequence is based on the lost of life which isaffected by fire, toxic and explosive. The calculation for one industrialarea simple model as case study was done. From the calculation result, theconsequences value of 0 - 25 persons/event; 26 - 50 persons/event; 51 - 75persons/event, 625 - 650 persons/event and the event probability value of1.10 -2 to 3.10 -8 are obtained. The optimization value of industrial areaaround can be determined by using combination of probability value andconsequences value. (author)

Evaluations and utilizations of risk importances

International Nuclear Information System (INIS)

Vesely, W.E.; Davis, T.C.

1985-08-01

This report presents approaches for utilizing Probabilistic Risk Analyses (PRA's) to determine risk importances. Risk importances are determined for design features, plant operations, and other factors that can affect risk. PRA's can be used to identify the importances of risk contributors or proposed changes to designs or operations. The objective of this report is to serve as a handbook and guide in evaluating and applying risk importances. The utilization of both qualitative risk importances and quantitative risk importances is described in this report. Qualitative risk importances are based on the logic models in the PRA, while quantitative risk importances are based on the quantitative results of the PRA. Both types of importances are among the most robust and meaningful information a PRA can provide. A wide variety of risk importance evaluations are described including evaluations of the importances of design changes, testing, maintenance, degrading environments, and aging. Specific utilizations are described in inspection and in reliability assurance programs, however the general approaches have widespread applicability. The role of personal computers and decision support programs in applying risk importance evaluations is also described

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.

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)

Relay chatter and operator response after a large earthquake: An improved PRA methodology with case studies

International Nuclear Information System (INIS)

Budnitz, R.J.; Lambert, H.E.; Hill, E.E.

1987-08-01

The purpose of this project has been to develop and demonstrate improvements in the PRA methodology used for analyzing earthquake-induced accidents at nuclear power reactors. Specifically, the project addresses methodological weaknesses in the PRA systems analysis used for studying post-earthquake relay chatter and for quantifying human response under high stress. An improved PRA methodology for relay-chatter analysis is developed, and its use is demonstrated through analysis of the Zion-1 and LaSalle-2 reactors as case studies. This demonstration analysis is intended to show that the methodology can be applied in actual cases, and the numerical values of core-damage frequency are not realistic. The analysis relies on SSMRP-based methodologies and data bases. For both Zion-1 and LaSalle-2, assuming that loss of offsite power (LOSP) occurs after a large earthquake and that there are no operator recovery actions, the analysis finds very many combinations (Boolean minimal cut sets) involving chatter of three or four relays and/or pressure switch contacts. The analysis finds that the number of min-cut-set combinations is so large that there is a very high likelihood (of the order of unity) that at least one combination will occur after earthquake-caused LOSP. This conclusion depends in detail on the fragility curves and response assumptions used for chatter. Core-damage frequencies are calculated, but they are probably pessimistic because assuming zero credit for operator recovery is pessimistic. The project has also developed an improved PRA methodology for quantifying operator error under high-stress conditions such as after a large earthquake. Single-operator and multiple-operator error rates are developed, and a case study involving an 8-step procedure (establishing feed-and-bleed in a PWR after an earthquake-initiated accident) is used to demonstrate the methodology

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

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)

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.)

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

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

Generic Pest Risk Analysis for Potato in Nepal

Directory of Open Access Journals (Sweden)

Baidya Nath Mahto

2017-05-01

Full Text Available Pest Risk Analysis (PRA is the process of evaluation for biological and economic evidences in order to determine whether a pest should be regulated under phyto-sanitary measures. The present mini review highlights the potential potato pathogen list recorded in Nepal harmful for potato production and productivity. At global scale altogether 135 potential quarantine pests (PQP for potato has been recorded, while in Nepal only 92 PQP were recorded. Out of those 52, 13 and 27 were fungi, bacteria and viruses respectively. Among the 92 PQP, 34, 30 and 13 were considered at high, medium and lower risk type pathogens for potato. There was no information available on other 15 PQP.

Application of Risk Assessment Tools in the Continuous Risk Management (CRM) Process

Science.gov (United States)

Ray, Paul S.

2002-01-01

Marshall Space Flight Center (MSFC) of the National Aeronautics and Space Administration (NASA) is currently implementing the Continuous Risk Management (CRM) Program developed by the Carnegie Mellon University and recommended by NASA as the Risk Management (RM) implementation approach. The four most frequently used risk assessment tools in the center are: (a) Failure Modes and Effects Analysis (FMEA), Hazard Analysis (HA), Fault Tree Analysis (FTA), and Probabilistic Risk Analysis (PRA). There are some guidelines for selecting the type of risk assessment tools during the project formulation phase of a project, but there is not enough guidance as to how to apply these tools in the Continuous Risk Management process (CRM). But the ways the safety and risk assessment tools are used make a significant difference in the effectiveness in the risk management function. Decisions regarding, what events are to be included in the analysis, to what level of details should the analysis be continued, make significant difference in the effectiveness of risk management program. Tools of risk analysis also depends on the phase of a project e.g. at the initial phase of a project, when not much data are available on hardware, standard FMEA cannot be applied; instead a functional FMEA may be appropriate. This study attempted to provide some directives to alleviate the difficulty in applying FTA, PRA, and FMEA in the CRM process. Hazard Analysis was not included in the scope of the study due to the short duration of the summer research project.

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)

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

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.)

Cardiovascular risk from water arsenic exposure in Vietnam: Application of systematic review and meta-regression analysis in chemical health risk assessment.

Science.gov (United States)

Phung, Dung; Connell, Des; Rutherford, Shannon; Chu, Cordia

2017-06-01

A systematic review (SR) and meta-analysis cannot provide the endpoint answer for a chemical risk assessment (CRA). The objective of this study was to apply SR and meta-regression (MR) analysis to address this limitation using a case study in cardiovascular risk from arsenic exposure in Vietnam. Published studies were searched from PubMed using the keywords of arsenic exposure and cardiovascular diseases (CVD). Random-effects meta-regression was applied to model the linear relationship between arsenic concentration in water and risk of CVD, and then the no-observable-adverse-effect level (NOAEL) were identified from the regression function. The probabilistic risk assessment (PRA) technique was applied to characterize risk of CVD due to arsenic exposure by estimating the overlapping coefficient between dose-response and exposure distribution curves. The risks were evaluated for groundwater, treated and drinking water. A total of 8 high quality studies for dose-response and 12 studies for exposure data were included for final analyses. The results of MR suggested a NOAEL of 50 μg/L and a guideline of 5 μg/L for arsenic in water which valued as a half of NOAEL and guidelines recommended from previous studies and authorities. The results of PRA indicated that the observed exposure level with exceeding CVD risk was 52% for groundwater, 24% for treated water, and 10% for drinking water in Vietnam, respectively. The study found that systematic review and meta-regression can be considered as an ideal method to chemical risk assessment due to its advantages to bring the answer for the endpoint question of a CRA. Copyright © 2017 Elsevier Ltd. All rights reserved.

Risk Management of NASA Projects

Science.gov (United States)

Sarper, Hueseyin

1997-01-01

Various NASA Langley Research Center and other center projects were attempted for analysis to obtain historical data comparing pre-phase A study and the final outcome for each project. This attempt, however, was abandoned once it became clear that very little documentation was available. Next, extensive literature search was conducted on the role of risk and reliability concepts in project management. Probabilistic risk assessment (PRA) techniques are being used with increasing regularity both in and outside of NASA. The value and the usage of PRA techniques were reviewed for large projects. It was found that both civilian and military branches of the space industry have traditionally refrained from using PRA, which was developed and expanded by nuclear industry. Although much has changed with the end of the cold war and the Challenger disaster, it was found that ingrained anti-PRA culture is hard to stop. Examples of skepticism against the use of risk management and assessment techniques were found both in the literature and in conversations with some technical staff. Program and project managers need to be convinced that the applicability and use of risk management and risk assessment techniques is much broader than just in the traditional safety-related areas of application. The time has come to begin to uniformly apply these techniques. The whole idea of risk-based system can maximize the 'return on investment' that the public demands. Also, it would be very useful if all project documents of NASA Langley Research Center, pre-phase A through final report, are carefully stored in a central repository preferably in electronic format.

Review of UCN 5,6 Fire PSA Model based on ANS Fire PRA Standard

International Nuclear Information System (INIS)

Yang, Joon Eon; Lee, Yoon Hwan

2006-12-01

Recently, under the de-regulation environment, nuclear industry has attempted various approaches to improve the economics of Nuclear Power Plants (NPP). This approach uses the fire risk and performance information to manage the resources effectively and efficiently that are used in the operation of NPP. In fire risk informed/performance-based decision/operation, fire 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 fire PSA. ANS (American Nuclear Society) has developed a guidance called 'ANS Fire PRA Methodology Standard'. However, in Korea, there have been no attempts to evaluate the quality of fire PSA model itself. Therefore, we cannot be sure about the quality of fire PSA whether or not the present fire PSA model can be used for the risk-informed applications such as mentioned above. We can say that the evaluation of fire PSA model quality is the basis for the fire risk informed/performance-based decision/operation. In this report, we have evaluated the quality of fire PSA model for Ulchin 5 and 6 units based on the ANS Fire PRA Standard. We, also, have derived what items are to be improved to upgrade the quality of fire PSA model and how it can be improved. This report can be used as the base of the fire risk informed/performance-based decision/operation work in Korea

Probabilistic risk assessment methodology for risk management and regulatory applications

International Nuclear Information System (INIS)

See Meng Wong; Kelly, D.L.; Riley, J.E.

1997-01-01

This paper discusses the development and potential applications of PRA methodology for risk management and regulatory applications in the U.S. nuclear industry. The new PRA methodology centers on the development of This paper discusses the time-dependent configuration risk profile for evaluating the effectiveness of operational risk management programs at U.S. nuclear power plants. Configuration-risk profiles have been used as risk-information tools for (1) a better understanding of the impact of daily operational activities on plant safety, and (2) proactive planning of operational activities to manage risk. Trial applications of the methodology were undertaken to demonstrate that configuration-risk profiles can be developed routinely, and can be useful for various industry and regulatory applications. Lessons learned include a better understanding of the issues and characteristics of PRA models available to industry, and identifying the attributes and pitfalls in the developement of risk profiles

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

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.

The approach to risk analysis in three industries

International Nuclear Information System (INIS)

Garrick, B.J.

1991-01-01

It is the purpose of this paper to review how risk and safety analysis is performed in the three major industries of nuclear power, space flight, and chemical and petroleum processes. The underlying reason for such a review is the belief that efficiencies and safety enhancements may result from a greater exchange of risk assessment technology between these industries. The thrust of this discussion related to the engineered systems involved in the three industries. The industries are very different. The chemical industry epitomizes the highly competitive private sector and its bottom-line emphasis; the nuclear power industry is unique by the degree to which it is regulated; and the space industry is essentially a government business just beginning to have commercial implications. Institutional differences are extreme; however, from a societal needs, and their safety implications have a far reaching impact on public opinion and support. In reviewing the risk and safety analysis activities, particular attention is given to the use of such quantitative approaches as probabilistic risk assessment (PRA) as it has evolved in the nuclear power industry

Integrated Reliability and Risk Analysis System (IRRAS)

International Nuclear Information System (INIS)

Russell, K.D.; McKay, M.K.; Sattison, M.B.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

1992-01-01

The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification. Version 1.0 of the IRRAS program was released in February of 1987. Since that time, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 4.0 and is the subject of this Reference Manual. Version 4.0 of IRRAS provides the same capabilities as Version 1.0 and adds a relational data base facility for managing the data, improved functionality, and improved algorithm performance

Risk Importance Measures in the Designand Operation of Nuclear Power Plants

Energy Technology Data Exchange (ETDEWEB)

Vrbanic I.; Samanta P.; Basic, I

2017-10-31

This monograph presents and discusses risk importance measures as quantified by the probabilistic risk assessment (PRA) models of nuclear power plants (NPPs) developed according to the current standards and practices. Usually, PRA tools calculate risk importance measures related to a single ?basic event? representing particular failure mode. This is, then, reflected in many current PRA applications. The monograph focuses on the concept of ?component-level? importance measures that take into account different failure modes of the component including common-cause failures (CCFs). In opening sections the roleof risk assessment in safety analysis of an NPP is introduced and discussion given of ?traditional?, mainly deterministic, design principles which have been established to assign a level of importance to a particular system, structure or component. This is followed by an overview of main risk importance measures for risk increase and risk decrease from current PRAs. Basic relations which exist among the measures are shown. Some of the current practical applications of risk importancemeasures from the field of NPP design, operation and regulation are discussed. The core of the monograph provides a discussion on theoreticalbackground and practical aspects of main risk importance measures at the level of ?component? as modeled in a PRA, starting from the simplest case, single basic event, and going toward more complexcases with multiple basic events and involvements in CCF groups. The intent is to express the component-level importance measures via theimportance measures and probabilities of the underlying single basic events, which are the inputs readily available from a PRA model andits results. Formulas are derived and discussed for some typical cases. The formulas and their results are demonstrated through some practicalexamples, done by means of a simplified PRA model developed in and run by RiskSpectrum? tool, which are presented in the appendices. The

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

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)

Analysis of risk-dominant sequences by MAAP3.0 for Kuosheng Nuclear Power Plant

International Nuclear Information System (INIS)

Lin, J.D.; Chieng, C.C.; Wang, T.K.; Hsiue, R.K.

1987-01-01

Kuosheng Nuclear Power Plant is the first operating model-6/Mark III boiling water reactor (BWR6/Mark III) in the world, and a probabilistic risk assessment (PRA) has been performed to determine the likely frequencies of core melt accidents and the magnitude, composition, and fraction of fission products released in these accidents. The final report of this PRA indicates that categories 8 and 15 are ranked No. 1 by risk index (the product of release frequency and release fraction) and release frequency, respectively. The dominant contributors of these two categories are frequent earthquakes and typhoons

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

The probabilistic risk analysis of external hazards of an interim storage for spent nuclear fuel in Olkiluoto

International Nuclear Information System (INIS)

Puukka, Tiia

2014-01-01

Due to natural disasters occurred in the world and the experiences perceived of the Fukushima nuclear accident, the particular knowledge of the role and influence of external hazards in the safety of interim storage of spent nuclear fuel has been emphasized. For that reason it is substantial that they are included in the probabilistic risk assessment (PRA) of the interim storage facility. This is also required by the Regulatory Guides issued by The Finnish Radiation and Nuclear Safety Authority STUK. To enhance safety culture and nuclear safety in Olkiluoto, The Finnish utility Teollisuuden Voima Oyj has recently completed an analysis of external natural (seismic events are studied as a separate analysis) and unintentional human-induced risks associated with the spent fuel pool cooling and decay heat removal systems as part of the full-scope PRA study for the interim storage of spent fuel (KPA store). The analysis had four goals to achieve: (1) to determine the definition of an initiating event in the context of the KPA store, (2) to identify all potential external hazards and hazard combinations, (3) to perform a qualitative screening analysis based on frequency-strength analysis and detailed plant responses analysis and (4) to model the hazards passed the screening analysis so that model can be used as a risk analysis tool in the risk informed decision making and operating procedures. The assessment carried out included the analysis of operation procedures of decay heat removal, the study of external hazards related initiating events included in the PRA of the OL1 and OL2 nuclear power plants and their dependencies on the initiating events of the KPA store. All external hazards related initiating events were modeled using fault tree linking method. The main result and conclusion of this study was that using the screening analysis, initiating events caused by external hazards that could lead to leakage of the spent fuel pools or that could pose a threat to the

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.

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

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.

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

Failure rate data for fusion safety and risk assessment

International Nuclear Information System (INIS)

Cadwallader, L.C.

1993-01-01

The Fusion Safety Program (FSP) at the Idaho National Engineering Laboratory (INEL) conducts safety research in materials, chemical reactions, safety analysis, risk assessment, and in component research and development to support existing magnetic fusion experiments and also to promote safety in the design of future experiments. One of the areas of safety research is applying probabilistic risk assessment (PRA) methods to fusion experiments. To apply PRA, we need a fusion-relevant radiological dose code and a component failure rate data base. This paper describes the FSP effort to develop a failure rate data base for fusion-specific components

An Example of Risk Informed Design

Science.gov (United States)

Banke, Rick; Grant, Warren; Wilson, Paul

2014-01-01

NASA Engineering requested a Probabilistic Risk Assessment (PRA) to compare the difference in the risk of Loss of Crew (LOC) and Loss of Mission (LOM) between different designs of a fluid assembly. They were concerned that the configuration favored by the design team was more susceptible to leakage than a second proposed design, but realized that a quantitative analysis to compare the risks between the two designs might strengthen their argument. The analysis showed that while the second design did help improve the probability of LOC, it did not help from a probability of LOM perspective. This drove the analysis team to propose a minor design change that would drive the probability of LOM down considerably. The analysis also demonstrated that there was another major risk driver that was not immediately obvious from a typical engineering study of the design and was therefore unexpected. None of the proposed alternatives were addressing this risk. This type of trade study demonstrates the importance of performing a PRA in order to completely understand a system's design. It allows managers to use risk as another one of the commodities (e.g., mass, cost, schedule, fault tolerance) that can be traded early in the design of a new system.

User's guide for PRISM (Plant Risk Status Information Management System) Arkansas Nuclear One-Unit 1: Volume 1, Program for inspectors

International Nuclear Information System (INIS)

Campbell, D.J.; Guthrie, V.H.; Kirchner, J.R.; Kirkman, J.Q.; Paula, H.M.; Ellison, B.C.; Dycus, F.M.; Farquharson, J.A.; Flanagan, G.F.

1988-03-01

This user's guide is a two-volume document designed to teach NRC inspectors and NRC regulators how to access probabilistic risk assessment information from the two Plant Risk Status Information Management System (PRISIM) programs developed for Arkansas Nuclear One -- Unit One (ANO-1). This document, Volume 1, describes how the PRA information available in Version 1.0 of PRISIM is useful for planning inspections. Using PRISIM, inspectors can quickly access PRA information and use that information to update risk analysis results, reflecting a plant's status at any particular time. Both volumes are stand-alone documents, and each volume presents several sample computer sessions designed to lead the user through a variety of PRISIM applications used to obtain PRA-related information for monitoring and controlling plant risk

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

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.)

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)

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

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...

Taking the Risk Out of Risk Assessment

Science.gov (United States)

2005-01-01

The ability to understand risks and have the right strategies in place when risky events occur is essential in the workplace. More and more organizations are being confronted with concerns over how to measure their risks or what kind of risks they can take when certain events transpire that could have a negative impact. NASA is one organization that faces these challenges on a daily basis, as effective risk management is critical to the success of its missions especially the Space Shuttle missions. On July 29, 1996, former NASA Administrator Daniel Goldin charged NASA s Office of Safety and Mission Assurance with developing a probabilistic risk assessment (PRA) tool to support decisions on the funding of Space Shuttle upgrades. When issuing the directive, Goldin said, "Since I came to NASA [in 1992], we've spent billions of dollars on Shuttle upgrades without knowing how much they improve safety. I want a tool to help base upgrade decisions on risk." Work on the PRA tool began immediately. The resulting prototype, the Quantitative Risk Assessment System (QRAS) Version 1.0, was jointly developed by NASA s Marshall Space Flight Center, its Office of Safety and Mission Assurance, and researchers at the University of Maryland. QRAS software automatically expands the reliability logic models of systems to evaluate the probability of highly detrimental outcomes occurring in complex systems that are subject to potential accident scenarios. Even in its earliest forms, QRAS was used to begin PRA modeling of the Space Shuttle. In parallel, the development of QRAS continued, with the goal of making it a world-class tool, one that was especially suited to NASA s unique needs. From the beginning, an important conceptual goal in the development of QRAS was for it to help bridge the gap between the professional risk analyst and the design engineer. In the past, only the professional risk analyst could perform, modify, use, and perhaps even adequately understand PRA. NASA wanted

Risk-based regulation: A regulatory perspective

International Nuclear Information System (INIS)

Scarborough, J.C.

1993-01-01

In the early development of regulations for nuclear power plants, risk was implicitly considered through qualitative assessments and engineering reliability principles and practices. Examples included worst case analysis, defense in depth, and the single failure criterion. However, the contributions of various systems, structures, components and operator actions to plant safety were not explicitly assessed since a methodology for this purpose had not been developed. As a consequence of the TMI accident, the use of more quantitative risk methodology and information in regulation such as probabilistic risk analysis (PRA) increased. The use of both qualitative and quantitative consideration of risk in regulation has been a set of regulations and regulatory guides and practices that ensure adequate protection of public health and safety. Presently, the development of PRA techniques has developed to the point that safety goals, expressed in terms of risk, have been established to help guide further regulatory decision making. This paper presents the personal opinions of the author as regards the use of risk today in nuclear power plant regulation, areas of further information needs, and necessary plans for moving toward a more systematic use of risk-based information in regulatory initiatives in the future

Methods development to evaluate the risk of upgrading to DCS: The human factor

Energy Technology Data Exchange (ETDEWEB)

Ostrom, L.T.; Wilhelmsen, C.A. [Idaho National Engineering Lab., Idaho Falls, ID (United States)

1995-04-01

The NRC recognizes that a more complete technical basis for understanding and regulating advanced digital technologies in commercial nuclear power plants is needed. A concern is that the introduction of digital safety systems may have an impact on risk. There is currently no standard methodology for measuring digital system reliability. A tool currently used to evaluate NPP risk in analog systems is the probabilistic risk assessment (PRA). The use of this tool to evaluate the digital system risk was considered to be a potential methodology for determining the risk. To test this hypothesis, it was decided to perform a limited PRA on a single dominant accident sequence. However, a review of existing human reliability analysis (HRA) methods showed that they were inadequate to analyze systems utilizing digital technology. A four step process was used to adapt existing HRA methodologies to digital environments and to develop new techniques. The HRA methods were then used to analyze an NPP that had undergone a backfit to digital technology in order to determine, as a first step, whether the methods were effective. The very small-break loss of coolant accident sequence was analyzed to determine whether the upgrade to the Eagle-21 process protection system had an effect on risk. The analysis of the very small-break LOCA documented in the Sequoyah PRA was used as the basis of the analysis. The analysis of the results of the HRA showed that the mean human error probabilities for the Eagle-21 PPS were slightly less than those for the analog system it replaced. One important observation from the analysis is that the operators have increased confidence steming from the better level of control provided by the digital system. The analysis of the PRA results, which included the human error component and the Eagle-21 PPS, disclosed that the reactor protection system had a higher failure rate than the analog system, although the difference was not statistically significant.

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

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

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.

Preliminary risk analysis applied to the handling of health-care waste

Directory of Open Access Journals (Sweden)

Carvalho S.M.L.

2002-01-01

Full Text Available Between 75% and 90% of the waste produced by health-care providers no risk or is "general" health-care waste, comparable to domestic waste. The remaining 10-25% of health-care waste is regarded as hazardous due to one or more of the following characteristics: it may contain infectious agents, sharps, toxic or hazardous chemicals or it may be radioactive. Infectious health-care waste, particularly sharps, has been responsible for most of the accidents reported in the literature. In this work the preliminary risks analysis (PRA technique was used to evaluate practices in the handling of infectious health-care waste. Currently the PRA technique is being used to identify and to evaluate the potential for hazard of the activities, products, and services from facilities and industries. The system studied was a health-care establishment which has handling practices for infectious waste. Thirty-six procedures related to segregation, containment, internal collection, and storage operation were analyzed. The severity of the consequences of the failure (risk that can occur from careless management of infectious health-care waste was classified into four categories: negligible, marginal, critical, and catastrophic. The results obtained in this study showed that events with critics consequences, about 80%, may occur during the implementation of the containment operation, suggesting the need to prioritize this operation. As a result of the methodology applied in this work, a flowchart the risk series was also obtained. In the flowchart the events that can occur as a consequence of a improper handling of infectious health-care waste, which can cause critical risks such as injuries from sharps and contamination (infection from pathogenic microorganisms, are shown.

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0. Volume 5, Systems Analysis and Risk Assessment (SARA) tutorial manual

International Nuclear Information System (INIS)

Sattison, M.B.; Russell, K.D.; Skinner, N.L.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs) primarily for nuclear power plants. This volume is the tutorial manual for the Systems Analysis and Risk Assessment (SARA) System Version 5.0, a microcomputer-based system used to analyze the safety issues of a open-quotes familyclose quotes [i.e., a power plant, a manufacturing facility, any facility on which a probabilistic risk assessment (PRA) might be performed]. A series of lessons is provided that guides the user through some basic steps common to most analyses performed with SARA. The example problems presented in the lessons build on one another, and in combination, lead the user through all aspects of SARA sensitivity analysis capabilities

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

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

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.

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)

Internal fire analysis screening methodology for the Salem Nuclear Generating Station

International Nuclear Information System (INIS)

Eide, S.; Bertucio, R.; Quilici, M.; Bearden, R.

1989-01-01

This paper reports on an internal fire analysis screening methodology that has been utilized for the Salem Nuclear Generating Station (SNGS) Probabilistic Risk Assessment (PRA). The methodology was first developed and applied in the Brunswick Steam Electric Plant (BSEP) PRA. The SNGS application includes several improvements and extensions to the original methodology. The SNGS approach differs significantly from traditional fire analysis methodologies by providing a much more detailed treatment of transient combustibles. This level of detail results in a model which is more usable for assisting in the management of fire risk at the plant

Comments of the PRA Senior Review Panel on the meeting held December 1--3, 1987

International Nuclear Information System (INIS)

Sharp, D.A.

1988-01-01

This memorandum records the minutes of the PRA Senior Review Panel meeting held at Savannah River Laboratory (SRL) on December 1--3, 1987, and the report on that meeting written subsequently by the panel members. The minutes are contained as Attachment 2 of this memorandum, and the report as Attachment 1. The Panel indicated two principal concerns in their report: (1) that insufficient emphasis is being placed on the reliability data development program, and (2) that excessive detail is being built into the fault trees. These concerns have been addressed in a subsequent meeting with the Panel, held March 2--4, 1988. In addition, the members have been provided with a program document (Reference 1) indicating the extent, the timing, and the limitations of the data analysis effort for the PRA

Risk assessment under deep uncertainty: A methodological comparison

International Nuclear Information System (INIS)

Shortridge, Julie; Aven, Terje; Guikema, Seth

2017-01-01

Probabilistic Risk Assessment (PRA) has proven to be an invaluable tool for evaluating risks in complex engineered systems. However, there is increasing concern that PRA may not be adequate in situations with little underlying knowledge to support probabilistic representation of uncertainties. As analysts and policy makers turn their attention to deeply uncertain hazards such as climate change, a number of alternatives to traditional PRA have been proposed. This paper systematically compares three diverse approaches for risk analysis under deep uncertainty (qualitative uncertainty factors, probability bounds, and robust decision making) in terms of their representation of uncertain quantities, analytical output, and implications for risk management. A simple example problem is used to highlight differences in the way that each method relates to the traditional risk assessment process and fundamental issues associated with risk assessment and description. We find that the implications for decision making are not necessarily consistent between approaches, and that differences in the representation of uncertain quantities and analytical output suggest contexts in which each method may be most appropriate. Finally, each methodology demonstrates how risk assessment can inform decision making in deeply uncertain contexts, informing more effective responses to risk problems characterized by deep uncertainty. - Highlights: • We compare three diverse approaches to risk assessment under deep uncertainty. • A simple example problem highlights differences in analytical process and results. • Results demonstrate how methodological choices can impact risk assessment results.

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.)

User's guide for PRISM (Plant Risk Status Information Management System) Arkansas Nuclear One-Unit 1: Volume 1, Program for inspectors

Energy Technology Data Exchange (ETDEWEB)

Campbell, D.J.; Guthrie, V.H.; Kirchner, J.R.; Kirkman, J.Q.; Paula, H.M.; Ellison, B.C.; Dycus, F.M.; Farquharson, J.A.; Flanagan, G.F.

1988-03-01

This user's guide is a two-volume document designed to teach NRC inspectors and NRC regulators how to access probabilistic risk assessment information from the two Plant Risk Status Information Management System (PRISIM) programs developed for Arkansas Nuclear One -- Unit One (ANO-1). This document, Volume 1, describes how the PRA information available in Version 1.0 of PRISIM is useful for planning inspections. Using PRISIM, inspectors can quickly access PRA information and use that information to update risk analysis results, reflecting a plant's status at any particular time. Both volumes are stand-alone documents, and each volume presents several sample computer sessions designed to lead the user through a variety of PRISIM applications used to obtain PRA-related information for monitoring and controlling plant risk.

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).

Bruce NGS B risk assessment (BBRA) peer review process

International Nuclear Information System (INIS)

Kaasalainen, S.; Crocker, W.P.; Webb, W.A.

2001-01-01

Risk-informed decision making is considered an effective approach to managing the risk of nuclear power plant operation in a competitive market. Hence, increased reliance on the station probabilistic risk assessments (PRAs) to provide risk perspective inputs is inevitable. With increased reliance on the PRAs it is imperative that PRAs have the characteristics necessary to provide the required information. Recognizing the increased requirements on nuclear power plant PRAs the nuclear industry in the United States has expended significant effort over the past few years defining the required characteristics of a PRA for various applications. More recently several owners groups have drafted guidelines for PRA certification and several U.S. utilities have had their PRAs certified. During the year 2000 Ontario Power Generation, Nuclear (OPG,N) subjected the PRA of one of its stations to the U.S. style certification process. The PRA selected for this process was the Bruce B Risk Assessment (BBRA). BBRA was chosen for this process since it is the first OPG, N PRA to be used for risk-informed applications. However, the strengths of the BBRA identified from the certification process and the lessons learned are also largely applicable to the other OPG, N plant PRAs due to the use of similar methods and tools

The prioritisation of a short list of alien plants for risk analysis within the framework of the Regulation (EU No. 1143/2014

Directory of Open Access Journals (Sweden)

Rob Tanner

2017-06-01

Full Text Available Thirty-seven alien plant species, pre-identified by horizon scanning exercises were prioritised for pest risk analysis (PRA using a modified version of the EPPO Prioritisation Process designed to be compliant with the EU Regulation 1143/2014. In Stage 1, species were categorised into one of four lists – a Residual List, EU List of Minor Concern, EU Observation List and the EU List of Invasive Alien Plants. Only those species included in the latter proceeded to the risk management stage where their priority for PRA was assessed. Due to medium or high spread potential coupled with high impacts twenty-two species were included in the EU List of Invasive Alien Plants and proceeded to Stage 2. Four species (Ambrosia trifida, Egeria densa, Fallopia baldschuanica and Oxalis pes-caprae were assigned to the EU Observation List due to moderate or low impacts. Albizia lebbeck, Clematis terniflora, Euonymus japonicus, Lonicera morrowii, Prunus campanulata and Rubus rosifolius were assigned to the residual list due to a current lack of information on impacts. Similarly, Cornus sericea and Hydrilla verticillata were assigned to the Residual List due to unclear taxonomy and uncertainty in native status, respectively. Chromolaena odorata, Cryptostegia grandiflora and Sphagneticola trilobata were assigned to the Residual List as it is unlikely they will establish in the Union under current climatic conditions. In the risk management stage, Euonymus fortunei, Ligustrum sinense and Lonicera maackii were considered a low priority for PRA as they do not exhibit invasive tendencies despite being widely cultivated in the EU over several decades. Nineteen species were identified as having a high priority for a PRA (Acacia dealbata, Ambrosia confertiflora, Andropogon virginicus, Cardiospermum grandiflorum, Celastrus orbiculatus, Cinnamomum camphora, Cortaderia jubata, Ehrharta calycina, Gymnocoronis spilanthoides, Hakea sericea, Humulus scandens, Hygrophila polysperma

Formalized Search Strategies for Human Risk Contributions

DEFF Research Database (Denmark)

Rasmussen, Jens; Pedersen, O. M.

For risk management, the results of a probabilistic risk analysis (PRA) as well as the underlying assumptions can be used as references in a closed-loop risk control; and the analyses of operational experiences as a means of feedback. In this context, the need for explicit definition...... risk contributions are described on the basis of general plant design features relevant for risk and accident analysis. With this background, search strategies for human risk contributions are treated: Under the designation "work analysis", procedures for the analysis of familiar, well trained, planned...... tasks are proposed. Strategies for identifying human risk contributions outside this category are outlined....

Risk assessment of a fusion-reactor fuel-processing system

International Nuclear Information System (INIS)

Bruske, S.Z.; Holland, D.F.

1983-07-01

The probabilistic risk assessment (PRA) methodology provides a means to systematically examine the potential for accidents that may result in a release of hazardous materials. This report presents the PRA for a typical fusion reactor fuel processing system. The system used in the analysis is based on the Tritium Systems Test Assembly, which is being operated at the Los Alamos National Laboratory. The results of the evaluation are presented in a probability-consequence plot that describes the probability of various accidental tritium release magnitudes

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

Real-time risk assessment of operational events

International Nuclear Information System (INIS)

Perryman, L.J.; Foster, N.A.S.; Nicholls, D.R.; Grobbelaar, J.F.

1996-01-01

Probabilistic risk assessment (PRA) has always been fundamental to the licensing process of Koeberg nuclear power station. Furthermore, over the past 8 years PRA has assisted in many areas of operation. One of these areas is the real-time assessment of abnormal operating events. Over the years, considerable experience has been gained in using PRA to improve plant safety and performance. This paper presents some of the insights obtained in using PRA in such a dynamic role and demonstrates that, by developing and using the plant-specific 'living' PRA, considerable safety and financial gains can be obtained. These insights specifically concern the prerequisites before optimal use of a plant-specific 'living' PRA can be made. Finally, examples are presented of occurrences when PRA was used to improve plant safety and performance. These examples serve to demonstrate the advantages that can be obtained if sufficient resources are placed at the disposal of the PRA team. (orig.)

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

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.

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

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

Structural reliability analysis and seismic risk assessment

International Nuclear Information System (INIS)

Hwang, H.; Reich, M.; Shinozuka, M.

1984-01-01

This paper presents a reliability analysis method for safety evaluation of nuclear structures. By utilizing this method, it is possible to estimate the limit state probability in the lifetime of structures and to generate analytically the fragility curves for PRA studies. The earthquake ground acceleration, in this approach, is represented by a segment of stationary Gaussian process with a zero mean and a Kanai-Tajimi Spectrum. All possible seismic hazard at a site represented by a hazard curve is also taken into consideration. Furthermore, the limit state of a structure is analytically defined and the corresponding limit state surface is then established. Finally, the fragility curve is generated and the limit state probability is evaluated. In this paper, using a realistic reinforced concrete containment as an example, results of the reliability analysis of the containment subjected to dead load, live load and ground earthquake acceleration are presented and a fragility curve for PRA studies is also constructed

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

Integrating risk management and safety culture in a framework for risk informed decision making

International Nuclear Information System (INIS)

Nelson, W.R.

2009-01-01

Operators and regulators of nuclear power plants agree on the importance of maintaining safety and controlling accident risks. Effective safety and risk management requires treatment of both technical and organizational components. Probabilistic Risk Assessment (PRA) provides tools for technical risk management. However, organizational factors are not treated in PRA, but are addressed using different approaches. To bring both components together, a framework of Risk Informed Decision Making (RIDM) is needed. The objective tree structure of the International Atomic Energy Agency (IAEA) is a promising approach to combine both elements. Effective collaboration involving regulatory and industry groups is needed to accomplish the integration. (author)

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.

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.

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

Systamatic approach to integration of a human reliability analysis into a NPP probabalistic risk assessment

International Nuclear Information System (INIS)

Fragola, J.R.

1984-01-01

This chapter describes the human reliability analysis tasks which were employed in the evaluation of the overall probability of an internal flood sequence and its consequences in terms of disabling vulnerable risk significant equipment. Topics considered include the problem familiarization process, the identification and classification of key human interactions, a human interaction review of potential initiators, a maintenance and operations review, human interaction identification, quantification model selection, the definition of operator-induced sequences, the quantification of specific human interactions, skill- and rule-based interactions, knowledge-based interactions, and the incorporation of human interaction-related events into the event tree structure. It is concluded that an integrated approach to the analysis of human interaction within the context of a Probabilistic Risk Assessment (PRA) is feasible

Statistically based uncertainty assessments in nuclear risk analysis

International Nuclear Information System (INIS)

Spencer, F.W.; Diegert, K.V.; Easterling, R.G.

1987-01-01

Over the last decade, the problems of estimation and uncertainty assessment in probabilistics risk assessment (PRAs) have been addressed in a variety of NRC and industry-sponsored projects. These problems have received attention because of a recognition that major uncertainties in risk estimation exist, which can be reduced by collecting more and better data and other information, and because of a recognition that better methods for assessing these uncertainties are needed. In particular, a clear understanding of the nature and magnitude of various sources of uncertainty is needed to facilitate descision-making on possible plant changes and research options. Recent PRAs have employed methods of probability propagation, sometimes involving the use of Bayes Theorem, and intended to formalize the use of ''engineering judgment'' or ''expert opinion.'' All sources, or feelings, of uncertainty are expressed probabilistically, so that uncertainty analysis becomes simply a matter of probability propagation. Alternatives to forcing a probabilistic framework at all stages of a PRA are a major concern in this paper, however

A new risk-informed design and regulatory process

International Nuclear Information System (INIS)

Apostolakis, George E.; Golay, Michael W.; Camp, Allen L.; Duran, Felicia A.; Finnicum, David; Ritterbusch, Stanley E.

2001-01-01

The overall purpose of the new approach, termed Risk-Informed Regulation, is to formulate a method of regulation that is logically consistent and devised so that both the reactor designer and regulator can work together in obtaining systems able to produce economical electricity safely. In this new system the traditional tools (deterministic and probabilistic analyses, tests and expert judgement) and treatments (defense-in-depth, conservatism) of safety regulation would still be employed, but the logic governing their use would be reversed from the current treatment. In the new treatment, probabilistic risk analysis (PRA) would be used as the paramount decision support tool, taking advantage of its ability to integrate all of the elements of system performance and to represent the uncertainties in the results. The latter is the most important reason for this choice, as the most difficult part of safety regulation is the treatment of uncertainties, not the assurance of expected performance. The scope of the PRA would be made as large as that of the reactor system, including all of its performance phenomena. The models and data of the PRA would be supported by deterministic analytical results, and data to the extent feasible. However, as in the current regulatory system, the models and data of the PRA would require being complemented by subjective judgements where the former were inadequate. All of these elements play important roles in the current decision-making structure; the main departure from current practice would be making all of these treatments explicit within the PRA, therefore, decreasing the frequency of sometimes arbitrary judgments. In the intended sense the PRA would be used as a vehicle for stating the beliefs of the designer and regulatory decision-maker; the foundation of their decisions. Thus, the PRA should be viewed as a Bayesian decision tool, and be used in order to take advantage of its capabilities in integration and inclusion of

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.

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;

An approach for risk management and regulatory applications

International Nuclear Information System (INIS)

Wong, See-Meng

2000-01-01

This paper discusses the development and potential applications of a PRA methodology for risk management and regulatory applications in the U.S. nuclear industry. The new PRA methodology centers on the development of time-dependent configuration risk profiles for evaluating the effectiveness of operational risk management programs at U.S. nuclear power plants. Configuration-risk profiles have been used as risk-information tools for (1) a better understanding of the impact of daily operational activities on plant safety and (2) proactive planning of operational activities to manage risk. Trial applications of the methodology were undertaken to demonstrate that configuration-risk profiles can be developed routinely, and can be useful for various industry and regulatory applications. Lessons learned include a better understanding of the issues and characteristics of PRA models available to industry, and identifying the attributes and pitfalls in the development of risk profiles. (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

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Rasmussen, Martin; Ulrich, Thomas; Groth, Katrina; Smith, Curtis

2016-01-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

Integration of Human Reliability Analysis Models into the Simulation-Based Framework for the Risk-Informed Safety Margin Characterization Toolkit

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rasmussen, Martin [Norwegian Univ. of Science and Technology, Trondheim (Norway). Social Research; Herberger, Sarah [Idaho National Lab. (INL), Idaho Falls, ID (United States); Ulrich, Thomas [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-06-01

This report presents an application of a computation-based human reliability analysis (HRA) framework called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER). HUNTER has been developed not as a standalone HRA method but rather as framework that ties together different HRA methods to model dynamic risk of human activities as part of an overall probabilistic risk assessment (PRA). While we have adopted particular methods to build an initial model, the HUNTER framework is meant to be intrinsically flexible to new pieces that achieve particular modeling goals. In the present report, the HUNTER implementation has the following goals: • Integration with a high fidelity thermal-hydraulic model capable of modeling nuclear power plant behaviors and transients • Consideration of a PRA context • Incorporation of a solid psychological basis for operator performance • Demonstration of a functional dynamic model of a plant upset condition and appropriate operator response This report outlines these efforts and presents the case study of a station blackout scenario to demonstrate the various modules developed to date under the HUNTER research umbrella.

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.

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

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

Component Degradation Susceptibilities As The Bases For Modeling Reactor Aging Risk

International Nuclear Information System (INIS)

Unwin, Stephen D.; Lowry, Peter P.; Toyooka, Michael Y.

2010-01-01

The extension of nuclear power plant operating licenses beyond 60 years in the United States will be necessary if we are to meet national energy needs while addressing the issues of carbon and climate. Characterizing the operating risks associated with aging reactors is problematic because the principal tool for risk-informed decision-making, Probabilistic Risk Assessment (PRA), is not ideally-suited to addressing aging systems. The components most likely to drive risk in an aging reactor - the passives - receive limited treatment in PRA, and furthermore, standard PRA methods are based on the assumption of stationary failure rates: a condition unlikely to be met in an aging system. A critical barrier to modeling passives aging on the wide scale required for a PRA is that there is seldom sufficient field data to populate parametric failure models, and nor is there the availability of practical physics models to predict out-year component reliability. The methodology described here circumvents some of these data and modeling needs by using materials degradation metrics, integrated with conventional PRA models, to produce risk importance measures for specific aging mechanisms and component types. We suggest that these measures have multiple applications, from the risk-screening of components to the prioritization of materials research.

Risk Management at NASA and Its Applicability to the Oil and Gas Industry

Science.gov (United States)

Kaplan, David

2018-01-01

NASA has a world-class capability for quantitatively assessing the risk of highly-complex, isolated engineering structures operated in extremely hostile environments. In particular, the International Space Station (ISS) represents a reasonable risk analog for High Pressure, High Temperature drilling and production operations on deepwater rigs. Through a long-term U.S. Government Interagency Agreement, BSEE has partnered with NASA to modify NASA's Probabilistic Risk Assessment (PRA) capabilities for application to deepwater drilling and production operations. The immediate focus of the activity will be to modify NASA PRA Procedure Guides and Methodology Documents to make them applicable to the Oil &Gas Industry. The next step will be for NASA to produce a PRA for a critical drilling system component, such as a Blowout Preventer (BOP). Subsequent activities will be for NASA and industry partners to jointly develop increasingly complex PRA's that analyze other critical drilling and production system components, including both hardware and human reliability. In the presentation, NASA will provide the objectives, schedule, and current status of its PRA activities for BSEE. Additionally, NASA has a Space Act Agreement with Anadarko Petroleum Corporation to develop a PRA for a generic 20K BOP. NASA will summarize some of the preliminary insights gained to date from that 20K BOP PRA as an example of the distinction between quantitative versus qualitative risk assessment.

A societal risk analysis model for nuclear power plants

International Nuclear Information System (INIS)

Klopp, George T.

2004-01-01

A review of the last decade and a half reveals that the nuclear power industry, world wide, has devoted increased attention to the concepts of reactor risk, probabilistic risk assessment (PRA), and cost benefit analyses. Millions of dollars have been spent by the industry and by regulatory agencies on studies of specific plants, research into severe accident behavior, and the development of national risk goals. In the United States, there is a major effort underway to evaluate each operating nuclear plant using PRA and the latest information on severe accident behavior. This effort constitutes a search for 'outliers' or vulnerabilities which may be profitably addressed by changes to plant design or operation. The question, then, immediately arises: How much is it reasonable to spend on this particular 'outlier?' The answer to this question, in each case, calls for some systematic vehicle for evaluating the worth of risk reduction. In turn, this calls for some means to look at all aspects of risk using a common yardstick or unit of measure. A review of past practices in such evaluations leads one directly to the classical cost benefit analyses which rarely use any guideline more comprehensive than the old $1000 per person-rem. The real costs of the TMI accident point to a need for a more realistic treatment. The BoPhal accident, the Chernobyl accident, and the Exxon Valdez accident highlight risk aspects previously not explored in detail and further support the postulate that a better method is mandated by history

An Approach to On-line Risk Assessment in NPP

International Nuclear Information System (INIS)

Simic, Z.; Mikulicic, V.; O'Brien, J.

1996-01-01

Probabilistic Risk Assessment (PRA) can provide safety status information for a plant during different configurations; additional effort is needed to do this in real time for on-line operation. This paper describes an approach to use PRA to achieve these goals. A Risk Assessment On-Line (RAOL) application was developed to monitor maintenance (on-line and planned) activities. RAOL is based on the results from a full-scope PRA, engineering/operational judgment and incorporates a user friendly program interface approach. Results from RAOL can be used by planners or operations to effectively manage the level of risk by controlling the actual plant configuration. (author)

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)

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.

PRAAGE-1988: An interactive IBM-PC code for aging analysis of NUREG-1150 systems

International Nuclear Information System (INIS)

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

1988-01-01

Probabilistic Risk Assessments (PRA) contain a great deal of information for estimating the risk of a nuclear power plant but do not consider aging. PRAAGE (PRA+AGE) is an interactive, IBM-PC code for processing PRA-developed system models using non-aged failure rate data in conjunction with user-supplied time-dependent nuclear plant experience component failure rate data to determine the effects of component aging on a system's reliability as well as providing the age-dependent importances of various generic components. This paper describes the structure, use and application of PRAAGE to the aging analysis of the Peach Bottom 2 RHR system in the LPCI and SDC modes of operation. 4 refs., 15 figs., 5 tabs

Research items regarding seismic residual risk evaluation

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

After learning the Fukushima Dai-ichi NPP severe accidents in 2011, the government investigation committee proposed the effective use of probabilistic safety assessment (PSA), and now it is required to establish new safety rules reflecting the results of probabilistic risk assessment (PRA) and proposed severe accident measures. Since the Seismic Design Guide has been revised on September 19, 2006, JNES has been discussing seismic PRA (Levels 1-3) methods to review licensees' residual risk assessment while preparing seismic PRA models. Meanwhile, new safety standards for light water reactors are to be issued and enforced on July 2013, which require the residual risk of tsunami, in addition to earthquakes, should be lowered as much as possible. The Fukushima accidents raised the problems related to risk assessment, e.g. approaches based on multi-hazard (earthquake and tsunami), multi-unit, multi-site, and equipment's common cause failure. This fiscal year, while performing seismic and/or tsunami PRA to work on these problems, JNES picked up the equipment whose failure greatly contribute to core damage, surveyed accident management measures on those equipment as well as effectiveness to reduce core damage probability. (author)

Data Analysis Approaches for the Risk-Informed Safety Margins Characterization Toolkit

International Nuclear Information System (INIS)

Mandelli, Diego; Alfonsi, Andrea; Maljovec, Daniel P.; Parisi, Carlo; Cogliati, Joshua J.; Talbot, Paul W.; Smith, Curtis L.; Rabiti, Cristian; Picoco, Claudia

2016-01-01

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, RELAP-7, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually called Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, ''extracting information'' means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework.

Data Analysis Approaches for the Risk-Informed Safety Margins Characterization Toolkit

Energy Technology Data Exchange (ETDEWEB)

Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Maljovec, Daniel P. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Parisi, Carlo [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Talbot, Paul W. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis L. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Picoco, Claudia [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-09-01

In the past decades, several numerical simulation codes have been employed to simulate accident dynamics (e.g., RELAP5-3D, RELAP-7, MELCOR, MAAP). In order to evaluate the impact of uncertainties into accident dynamics, several stochastic methodologies have been coupled with these codes. These stochastic methods range from classical Monte-Carlo and Latin Hypercube sampling to stochastic polynomial methods. Similar approaches have been introduced into the risk and safety community where stochastic methods (such as RAVEN, ADAPT, MCDET, ADS) have been coupled with safety analysis codes in order to evaluate the safety impact of timing and sequencing of events. These approaches are usually called Dynamic PRA or simulation-based PRA methods. These uncertainties and safety methods usually generate a large number of simulation runs (database storage may be on the order of gigabytes or higher). The scope of this paper is to present a broad overview of methods and algorithms that can be used to analyze and extract information from large data sets containing time dependent data. In this context, “extracting information” means constructing input-output correlations, finding commonalities, and identifying outliers. Some of the algorithms presented here have been developed or are under development within the RAVEN statistical framework.

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

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)

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)

Risk evaluation of medical and industrial radiation devices

International Nuclear Information System (INIS)

Jones, E.D.; Cunningham, R.E.; Rathbun, P.A.

1994-03-01

In 1991, the NRC, Division of Industrial and Medical Nuclear Safety, began a program to evaluate the use of probabilistic risk assessment (PRA) in regulating medical devices. This program represents an initial step in an overall plant to evaluate the use of PRA in regulating the use of nuclear by-product materials. The NRC envisioned that the use of risk analysis techniques could assist staff in ensuring that the regulatory approach was standardized, understandable, and effective. Traditional methods of assessing risk in nuclear power plants may be inappropriate to use in assessing the use of by-product devices. The approaches used in assessing nuclear reactor risks are equipment-oriented. Secondary attention is paid to the human component, for the most part after critical system failure events have been identified. This paper describes the risk methodology developed by Lawrence Livermore National Laboratory (LLNL), initially intended to assess risks associated with the use of the Gamma Knife, a gamma stereotactic radiosurgical device. For relatively new medical devices such as the Gamma Knife, the challenge is to perform a risk analysis with very little quantitative data but with an important human factor component. The method described below provides a basic approach for identifying the most likely risk contributors and evaluating their relative importance. The risk analysis approach developed for the Gamma Knife and described in this paper should be applicable to a broader class of devices in which the human interaction with the device is a prominent factor. In this sense, the method could be a prototypical model of nuclear medical or industrial device risk analysis

Development of a quantitative risk standard

International Nuclear Information System (INIS)

Temme, M.I.

1982-01-01

IEEE Working Group SC-5.4 is developing a quantitative risk standard for LWR plant design and operation. The paper describes the Working Group's conclusions on significant issues, including the scope of the standard, the need to define the process (i.e., PRA calculation) for meeting risk criteria, the need for PRA quality requirements and the importance of distinguishing standards from goals. The paper also describes the Working Group's approach to writing this standard

A scenario-based procedure for seismic risk analysis

International Nuclear Information System (INIS)

Kluegel, J.-U.; Mualchin, L.; Panza, G.F.

2006-12-01

A new methodology for seismic risk analysis based on probabilistic interpretation of deterministic or scenario-based hazard analysis, in full compliance with the likelihood principle and therefore meeting the requirements of modern risk analysis, has been developed. The proposed methodology can easily be adjusted to deliver its output in a format required for safety analysts and civil engineers. The scenario-based approach allows the incorporation of all available information collected in a geological, seismotectonic and geotechnical database of the site of interest as well as advanced physical modelling techniques to provide a reliable and robust deterministic design basis for civil infrastructures. The robustness of this approach is of special importance for critical infrastructures. At the same time a scenario-based seismic hazard analysis allows the development of the required input for probabilistic risk assessment (PRA) as required by safety analysts and insurance companies. The scenario-based approach removes the ambiguity in the results of probabilistic seismic hazard analysis (PSHA) which relies on the projections of Gutenberg-Richter (G-R) equation. The problems in the validity of G-R projections, because of incomplete to total absence of data for making the projections, are still unresolved. Consequently, the information from G-R must not be used in decisions for design of critical structures or critical elements in a structure. The scenario-based methodology is strictly based on observable facts and data and complemented by physical modelling techniques, which can be submitted to a formalised validation process. By means of sensitivity analysis, knowledge gaps related to lack of data can be dealt with easily, due to the limited amount of scenarios to be investigated. The proposed seismic risk analysis can be used with confidence for planning, insurance and engineering applications. (author)

A process for risk-focused maintenance

International Nuclear Information System (INIS)

Lofgren, E.V.; Cooper, S.E.; Kurth, R.E.; Phillips, L.B.

1991-03-01

This report presents a process for focusing maintenance resources on components that enable nuclear plant systems to perform their essential functions and on components whose failure may initiate challenges to safety systems, so as to have the greatest impact in decreasing risk. The process provides criteria, based on risk, for deciding which components are critical to risk and determining what maintenance activities are required to ensure reliable operation of those risk-critical components. Two approaches are provided for selection of risk-critical components. One approach uses the results of a Probabilistic Risk Assessment (PRA); the other is based on the methodology developed for this report, which has a basis in PRA although it does not use the results of a PRA study. Following identification of risk-critical components, both approaches use a single methodology for determining what maintenance activities are required to ensure reliable operation of the identified components. The report also provides demonstrations of application of the two approaches to selection of risk-critical components and demonstrations of application of the methodology for determining what maintenance activities are required to an active standby safety system, a normally operating system, and passive components. 5 refs., 11 figs., 1 tab

A process for risk-focused maintenance

Energy Technology Data Exchange (ETDEWEB)

Lofgren, E.V.; Cooper, S.E.; Kurth, R.E.; Phillips, L.B. (Science Applications International Corp., McLean, VA (USA))

1991-03-01

This report presents a process for focusing maintenance resources on components that enable nuclear plant systems to perform their essential functions and on components whose failure may initiate challenges to safety systems, so as to have the greatest impact in decreasing risk. The process provides criteria, based on risk, for deciding which components are critical to risk and determining what maintenance activities are required to ensure reliable operation of those risk-critical components. Two approaches are provided for selection of risk-critical components. One approach uses the results of a Probabilistic Risk Assessment (PRA); the other is based on the methodology developed for this report, which has a basis in PRA although it does not use the results of a PRA study. Following identification of risk-critical components, both approaches use a single methodology for determining what maintenance activities are required to ensure reliable operation of the identified components. The report also provides demonstrations of application of the two approaches to selection of risk-critical components and demonstrations of application of the methodology for determining what maintenance activities are required to an active standby safety system, a normally operating system, and passive components. 5 refs., 11 figs., 1 tab.

Integrated Reliability and Risk Analysis System (IRRAS), Version 2.5: Reference manual

International Nuclear Information System (INIS)

Russell, K.D.; McKay, M.K.; Sattison, M.B.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

1991-03-01

The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification. Version 1.0 of the IRRAS program was released in February of 1987. Since that time, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 2.5 and is the subject of this Reference Manual. Version 2.5 of IRRAS provides the same capabilities as Version 1.0 and adds a relational data base facility for managing the data, improved functionality, and improved algorithm performance. 7 refs., 348 figs

Review of UCN 3,4 PSA model based on NEI PRA peer review process guidance, rev.0

International Nuclear Information System (INIS)

Yang, Joon Eon; Kang, D. I.; Kim, K. Y.; Lee, Y. H.; Jang, S. C.; Ha, J. J.; Han, S. H.; Han, S. J.; Hwang, M. J.

2003-05-01

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 for Ulchin 3 and 4 units based on the NEI guidance. We, also, have derived what items are to be improved to upgrade the quality of PSA model and how it can be improved. This report can be used as the base of RIPBO work in Korea. The review result based on ASME Standard is published as the separated technical report of KAERI

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

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

IRRAS, Integrated Reliability and Risk Analysis System for PC

International Nuclear Information System (INIS)

Russell, K.D.

1995-01-01

1 - Description of program or function: IRRAS4.16 is a program developed for the purpose of performing those functions necessary to create and analyze a complete Probabilistic Risk Assessment (PRA). This program includes functions to allow the user to create event trees and fault trees, to define accident sequences and basic event failure data, to solve system and accident sequence fault trees, to quantify cut sets, and to perform uncertainty analysis on the results. Also included in this program are features to allow the analyst to generate reports and displays that can be used to document the results of an analysis. Since this software is a very detailed technical tool, the user of this program should be familiar with PRA concepts and the methods used to perform these analyses. 2 - Method of solution: IRRAS4.16 is written entirely in MODULA-2 and uses an integrated commercial graphics package to interactively construct and edit fault trees. The fault tree solving methods used are industry recognized top down algorithms. For quantification, the program uses standard methods to propagate the failure information through the generated cut sets. 3 - Restrictions on the complexity of the problem: Due to the complexity of and the variety of ways a fault tree can be defined it is difficult to define limits on the complexity of the problem solved by this software. It is, however, capable of solving a substantial fault tree due to efficient methods. At this time, the software can efficiently solve problems as large as other software currently used on mainframe computers. Does not include source code

Eliciting and Combining Decision Criteria Using a Limited Palette of Utility Functions and Uncertainty Distributions: Illustrated by Application to Pest Risk Analysis.

Science.gov (United States)

Holt, Johnson; Leach, Adrian W; Schrader, Gritta; Petter, Françoise; MacLeod, Alan; van der Gaag, Dirk Jan; Baker, Richard H A; Mumford, John D

2014-01-01

Utility functions in the form of tables or matrices have often been used to combine discretely rated decision-making criteria. Matrix elements are usually specified individually, so no one rule or principle can be easily stated for the utility function as a whole. A series of five matrices are presented that aggregate criteria two at a time using simple rules that express a varying degree of constraint of the lower rating over the higher. A further nine possible matrices were obtained by using a different rule either side of the main axis of the matrix to describe situations where the criteria have a differential influence on the outcome. Uncertainties in the criteria are represented by three alternative frequency distributions from which the assessors select the most appropriate. The output of the utility function is a distribution of rating frequencies that is dependent on the distributions of the input criteria. In pest risk analysis (PRA), seven of these utility functions were required to mimic the logic by which assessors for the European and Mediterranean Plant Protection Organization arrive at an overall rating of pest risk. The framework enables the development of PRAs that are consistent and easy to understand, criticize, compare, and change. When tested in workshops, PRA practitioners thought that the approach accorded with both the logic and the level of resolution that they used in the risk assessments. © 2013 Society for Risk Analysis.

Top event prevention analysis - a deterministic use of PRA

International Nuclear Information System (INIS)

Blanchard, D.P.; Worrell, R.B.

1995-01-01

Risk importance measures are popular for many applications of probabilistic analysis. Inherent in the derivation of risk importance measures are implicit assumptions that those using these numerical results should be aware of in their decision making. These assumptions and potential limitations include the following: (1) The risk importance measures are derived for a single event at a time and are therefore valid only if all other event probabilities are unchanged at their current values. (2) The results for which risk importance measures are derived may not be complete for reasons such as truncation

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.)

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.

Probabilistic Risk Analysis and Fault Trees as Tools in Improving the Delineation of Wellhead Protection Areas: An Initial Discussion

Science.gov (United States)

Rodak, C. M.; Silliman, S. E.

2010-12-01

Delineation of a wellhead protection area (WHPA) is a critical component of managing / protecting the aquifer(s) supplying potable water to a public water-supply well. While a number of previous authors have addressed questions related to uncertainties in advective capture zones, methods for assessing WHPAs in the presence of uncertainty in the chemistry of groundwater contaminants, the relationship between land-use and contaminant sources, and the impact on health risk within the receiving population are more limited. Probabilistic risk analysis (PRA) combined with fault trees (FT) addresses this latter challenge by providing a structure whereby four key WHPA issues may be addressed: (i) uncertainty in land-use practices and chemical release, (ii) uncertainty in groundwater flow, (iii) variability in natural attenuation properties (and/or remediation) of the contaminants, and (iv) estimated health risk from contaminant arrival at a well. The potential utility of PRA-FT in this application is considered through a simplified case study involving management decisions related both to regional land use planning and local land-use zoning regulation. An application-specific fault tree is constructed to visualize and identify the events required for health risk failure at the well and a Monte Carlo approach is used to create multiple realizations of groundwater flow and chemical transport to a well in a model of a simple, unconfined aquifer. Model parameters allowed to vary during this simplified case study include hydraulic conductivity, probability of a chemical spill (related to land use variation in space), and natural attenuation through variation in rate of decay of the contaminant. Numerical results are interpreted in association with multiple land-use management scenarios as well as multiple cancer risk assumptions regarding the contaminant arriving at the well. This case study shows significant variability of health risk at the well, however general trends were

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)

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.)

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)

Assessment and Control of Spacecraft Charging Risks on the International Space Station

Science.gov (United States)

Koontz, Steve; Valentine, Mark; Keeping, Thomas; Edeen, Marybeth; Spetch, William; Dalton, Penni

2004-01-01

The International Space Station (ISS) operates in the F2 region of Earth's ionosphere, orbiting at altitudes ranging from 350 to 450 km at an inclination of 51.6 degrees. The relatively dense, cool F2 ionospheric plasma suppresses surface charging processes much of the time, and the flux of relativistic electrons is low enough to preclude deep dielectric charging processes. The most important spacecraft charging processes in the ISS orbital environment are: 1) ISS electrical power system interactions with the F2 plasma, 2) magnetic induction processes resulting from flight through the geomagnetic field and, 3) charging processes that result from interaction with auroral electrons at high latitude. Recently, the continuing review and evaluation of putative ISS charging hazards required by the ISS Program Office revealed that ISS charging could produce an electrical shock hazard to the ISS crew during extravehicular activity (EVA). ISS charging risks are being evaluated in an ongoing measurement and analysis campaign. The results of ISS charging measurements are combined with a recently developed model of ISS charging (the Plasma Interaction Model) and an exhaustive analysis of historical ionospheric variability data (ISS Ionospheric Specification) to evaluate ISS charging risks using Probabilistic Risk Assessment (PRA) methods. The PRA combines estimates of the frequency of occurrence and severity of the charging hazards with estimates of the reliability of various hazard controls systems, as required by NASA s safety and risk management programs, to enable design and selection of a hazard control approach that minimizes overall programmatic and personnel risk. The PRA provides a quantitative methodology for incorporating the results of the ISS charging measurement and analysis campaigns into the necessary hazard reports, EVA procedures, and ISS flight rules required for operating ISS in a safe and productive manner.

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.

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

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

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)

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)

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

International Nuclear Information System (INIS)

Muta, Hitoshi; Muramatsu, Ken

2017-01-01

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

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

Development and trial application of risk information

Energy Technology Data Exchange (ETDEWEB)

NONE

2013-08-15

JNES has been doing various activities to stimulate the introduction of Risk Informed Regulation (RIR) to the safety regulation of nuclear power plants (NPPs) in Japan. Some applications are already incorporated, such as the regulatory review of Maintenance Programs and Safety Significance Evaluation for Inspection Findings. In consideration with the experience of the accident in Fukushima Daiichi Nuclear Power Station, JNES addressed development of regulatory guidelines, evaluation of the current condition of Fukushima Daiichi Nuclear Power Station, evaluation of effectiveness of severe accident management measures with the backgrounds of insights and experiences on probabilistic risk assessment (PRA) and RIR. Especially, the experiences were applied to the development of the methodologies for evaluation of effectiveness of severe accident managements. As for inspection and operation of NPPs, JNES enhanced the PRA scope applied to the importance analysis for Maintenance Program, SDP and RI-ISI in consideration with the insights of RIR in Japan and other countries. (author)

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.

Integrated Reliability and Risk Analysis System (IRRAS) Version 2.0 user's guide

International Nuclear Information System (INIS)

Russell, K.D.; Sattison, M.B.; Rasmuson, D.M.

1990-06-01

The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the user the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification. Also provided in the system is an integrated full-screen editor for use when interfacing with remote mainframe computer systems. Version 1.0 of the IRRAS program was released in February of 1987. Since that time, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 2.0 and is the subject of this user's guide. Version 2.0 of IRRAS provides all of the same capabilities as Version 1.0 and adds a relational data base facility for managing the data, improved functionality, and improved algorithm performance. 9 refs., 292 figs., 4 tabs

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.

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)

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 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

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

Role of frameworks, models, data, and judgment in human reliability analysis

Energy Technology Data Exchange (ETDEWEB)

Hannaman, G W

1986-05-01

Many advancements in the methods for treating human interactions in PRA studies have occurred in the last decade. These advancements appear to increase the capability of PRAs to extend beyond just the assessment of the human's importance to safety. However, variations in the application of these advanced models, data, and judgements in recent PRAs make quantitative comparisons among studies extremely difficult. This uncertainty in the analysis diminishes the usefulness of the PRA study for upgrading procedures, enhancing traning, simulator design, technical specification guidance, and for aid in designing the man-machine interface. Hence, there is a need for a framework to guide analysts in incorporating human interactions into the PRA systems analyses so that future users of a PRA study will have a clear understanding of the approaches, models, data, and assumptions which were employed in the initial study. This paper describes the role of the systematic human action reliability procedure (SHARP) in providing a road map through the complex terrain of human reliability that promises to improve the reproducibility of such analysis in the areas of selecting the models, data, representations, and assumptions. Also described is the role that a human cognitive reliability model can have in collecting data from simulators and helping analysts assign human reliability parameters in a PRA study. Use of these systematic approaches to perform or upgrade existing PRAs promises to make PRA studies more useful as risk management tools.

Efficacy of risk stratification in tailoring immunosuppression regimens in kidney transplant patients at the national kidney and transplant institute.

Science.gov (United States)

Ledesma-Gumba, M A; Danguilan, R A; Casasola, C C; Ona, E T

2008-09-01

To evaluate the efficacy of tailored immunosuppressive regimens prescribed according to a risk stratification scoring system based on the number of HLA mismatches, donor source, panel-reactive antibodies (PRA), and repeat transplant. Patients in a retrospective cohort of 329 kidney transplantations performed from October 2004 to December 2005 were assigned scores of 0, 2, 4, or 6 with higher scores for > or =1 HLA mismatches, PRA > 10%, repeat transplant, and unrelated or deceased donor. Added scores of or = 6 denoted high risk including a CNI-based regimen with an interleukin-2 receptor antibody. The efficacy analysis compared the incidences of biopsy-proven acute rejection episodes (BPAR) at 1 year. Only 227 (69%) of 329 patients had a complete data set and 84 were excluded because they did not follow the prescribed protocol, yielding 113 low- and 30 high-risk patients in the final population. Low-risk patients had a mean PRA of 5.4%, living related donors in 68%, and primary transplants. High-risk patients had a mean PRA of 18.8% (range = 10%-97%), living nonrelated donors in 84%, four deceased donors, and four repeat transplants. The overall 1-year incidence of BPAR was 5.7%. No significant difference (P = .081) was observed in 1-year BPAR between the low- (4.5%) and high-risk (9.8%) groups. Likewise, no significant difference in the 1-year mean serum creatinine was observed according to the CNI. The mean creatinine was 1.12 for cyclosporine and 1.38 for tacrolimus treatment (P = .06) in the low-risk group and 1.08 for cyclosporine and 1.2 for tacrolimus (P = .61) in the high-risk cohort. There was no significant difference in acute rejection rates between the immunologically low- or high-risk patients using tailored immunosuppression, which was effective to minimize its occurrence with good renal function at 1 year.

Overview of seismic margin insights gained from seismic PRA results

International Nuclear Information System (INIS)

Kennedy, R.P.; Sues, R.H.; Campbell, R.D.

1986-01-01

This paper presents the findings of a study conducted under NRC and EPRI sponsorship in which published seismic PRAs were reviewed in order to gain insight to the seismic margins inherent in existing nuclear plants. The approach taken was to examine the fragilities of those components which have been found to be dominant contributors to seismic risk at plants in low-to-moderate seismic regions (SSE levels between 0.12g and 0.25g). It is concluded that there is significant margin inherent in the capacity of most critical components above the plant design basis. For ground motions less than about 0.3g, the predominant sources of seismic risk are loss of offsite power coupled with random failure of the emergency diesels, non-recoverable circuit breaker trip due to relay chatter, unanchored equipment, unreinforced non-load bearing block walls, vertical water storage tanks, systems interactions and possibly soil liquefaction. Recommendations as to which components should be reviewed in seismic margin studies for margin earthquakes less than 0.3g, between 0.3g and 0.5g, and greater than 0.5g, developed by the NRC expert panel on the quantification of seismic margins (based on the review of past PRA data, earthquake experience data, and their own personal experience) are presented

Risk evaluation of accident management strategies

International Nuclear Information System (INIS)

Dingman, S.; Camp, A.

1992-01-01

The use of Probabilistic Risk Assessment (PRA) methods to evaluate accident management strategies in nuclear power plants discussed in this paper. The PRA framework allows an integrated evaluation to be performed to give the full implications of a particular strategy. The methodology is demonstrated for a particular accident management strategy, intentional depressurization of the reactor coolant system to avoid containment pressurization during the ejection of molten debris at vessel breach

Risk-based management system development for the Advanced Test Reactor

International Nuclear Information System (INIS)

Davis, M.L.; Eide, S.A.

1990-01-01

A Risk-Based Management System (RBMS) is being developed to facilitate the use of the Advanced Test Reactor (ATR) probabilistic risk assessment to support ATR operation. Most ATR RBMS questions can best be answered using the System Analysis and Risk Assessment System (SARA) developed at the Idaho National Engineering Laboratory. However, some applications may require employment of the other four codes used to develop and report the PRA. These four codes include the Integrated Reliability and Risk Analysis System (IRRAS), SETS, ETA-II, and the Nuclear Computerized Library for Assessing Reactor Reliability (NUCLARR). The ATR RBMS will evolve over three years, and will include the results of the Level 3 and external events analysis

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

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

Treatment of complementary events in event trees in constructing linked fault trees for level 1 and level 2 PRA

International Nuclear Information System (INIS)

Jo, Y. G.

2008-01-01

Complementary events in the event trees for a PRA model should be treated properly in order to evaluate plant risk correctly. In this study, the characteristics of the following three different cut-set generation methods were investigated first in order to find the best practical way for treating complementary events: 1) exact method which treats complementary events logically, 2) no-delete term method which does not treat complementary events at all, and 3) delete term method which treats complementary events by deleting nonsense cut-sets which are generated as a result of ignoring complementary events. Then, practical methods for treating complementary events in constructing linked fault trees for level 1 and level 2 PRA in EPRI R and R workstation software environment, where CAFTA is the fault tree editor and FORTE is the cut-set engine, were suggested and demonstrated. The suggested methods deal with the following selected four typical cases: Case 1: an event tree event (E) is represented by a fault tree gate whose inputs consist of only fault tree gates, Case 2: E is represented by a single basic event, Case 3: E is represented by an OR fault tree gate which has a single basic event and a fault tree gate as inputs, and Case 4: E is represented by an AND fault tree gate which has a single basic event and a fault tree gate as inputs. In the suggested methods, first the high level logic structures of event tree events are examined and restructured, if needed. Then, the delete term method, the exact method, and the combination of the two methods are applied to Case 1, Case 2, and Cases 3 and 4, respectively. Also, it is recommended to treat complementary events, using the suggested methods, before level 1 and level 2 PRA fault trees are coupled. It should be noted that the selected four typical cases may not cover all different cases encountered in level 1 and level 2 PRA modeling. However, a process similar to the one suggested in this study may be used to find

Methods for external event screening quantification: Risk Methods Integration and Evaluation Program (RMIEP) methods development

International Nuclear Information System (INIS)

Ravindra, M.K.; Banon, H.

1992-07-01

In this report, the scoping quantification procedures for external events in probabilistic risk assessments of nuclear power plants are described. External event analysis in a PRA has three important goals; (1) the analysis should be complete in that all events are considered; (2) by following some selected screening criteria, the more significant events are identified for detailed analysis; (3) the selected events are analyzed in depth by taking into account the unique features of the events: hazard, fragility of structures and equipment, external-event initiated accident sequences, etc. Based on the above goals, external event analysis may be considered as a three-stage process: Stage I: Identification and Initial Screening of External Events; Stage II: Bounding Analysis; Stage III: Detailed Risk Analysis. In the present report, first, a review of published PRAs is given to focus on the significance and treatment of external events in full-scope PRAs. Except for seismic, flooding, fire, and extreme wind events, the contributions of other external events to plant risk have been found to be negligible. Second, scoping methods for external events not covered in detail in the NRC's PRA Procedures Guide are provided. For this purpose, bounding analyses for transportation accidents, extreme winds and tornadoes, aircraft impacts, turbine missiles, and chemical release are described

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

Beyond informed choice: Prenatal risk assessment, decision-making and trust

Directory of Open Access Journals (Sweden)

Nete Schwennesen

2008-05-01

Full Text Available In 2004 prenatal risk assessment (PRA was implemented as a routine offer to all pregnant women in Denmark. It was argued that primarily the new programme would give all pregnant women an informed choice about whether to undergo prenatal testing. On the basis of ethnographic fieldwork in an ultrasound clinic in Denmark and interviews with pregnant women and their partners, we call into question the assumption underlying the new guidelines that more choice and more objective information is a source of empowerment and control. We focus on one couple's experience of PRA. This case makes it evident how supposed choices in the context of PRA may not be experienced as such. Rather, they are experienced as complicated processes of meaning-making in the relational space between the clinical setting, professional authority and the social life of the couples. PRA users are reluctant to make choices and abandon health professionals as authoritative experts in the face of complex risk knowledge. When assumptions about autonomy and self-determination are inscribed into the social practice of PRA, authority is transferred to the couple undergoing PRA and a new configuration of responsibility evolves between the couple and their relationship to the foetus. It is argued that al-though the new programme of prenatal testing in Denmark presents itself in opposition to quasi-eugenic and paternalistic forms of governing couples' decisions it represents another form of government that works through the notion of choice. An ethics of a shared responsibility of PRA and its outcome would be more in agreement with how decisions are actually made.http://dx.doi.org/10.5324/eip.v2i1.1687

Some insights from fire risk analysis of US nuclear power plants

International Nuclear Information System (INIS)

Kazarians, M.; Lambright, J.A.; Frank, M.V.

1998-01-01

Fire risk analysis has been conducted for a significant portion of the nuclear power plants in the U.S. using either Probabilistic Risk Assessments (PRAs) or FIVE or a combination of the two methodologies. Practically all fire risk studies have used step-wise, screening approach. To establish the contents of a compartment, the cable routing information collected for Appendix R compliance have been used in practically all risk studies. In several cases, the analysts have gone beyond the Appendix R and have obtained the routing of additional cables. For fire impact analysis typically an existing PRA model is used. For fire frequencies, typically, a generic data base is used. Fire scenarios are identified in varying levels of detail. The most common approach, in the early stages of screening, is based on the assumption that given a fire, the entire contents of the compartment are lost. Less conservative scenarios are introduced at later stages of the analysis which may include fire propagation patterns, fires localized to an item. and suppression of the fire before critical damage. For fire propagation and damage analysis, a large number of studies have used FIVE and many have used COMPBRN. For detection and suppression analysis, the generic suppression system unavailabilities given in FIVE have been used. The total core damage frequencies typically range between 1x10 -6 to 1x10 -4 per year. Control rooms and cable spreading rooms are the two most common areas found to be significant contributors to fire risk. Other areas are mainly from the Auxiliary Building (in the case of PWRs) and Reactor Building (in the case of BWRs). Only in one case, the main contributor to fire is the turbine building, which included several safety related equipment and cables. (author)

Risk Analysis of Fukushima Accident using MACCS2

Energy Technology Data Exchange (ETDEWEB)

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

2014-05-15

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

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

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

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

Evaluation of safety issues on newly regulated nuclear power plant by tsunami-level 1 PRA

International Nuclear Information System (INIS)

Tsuji, Yutaro; Miwa, Shuichiro; Mori, Michitsugu

2014-01-01

The tsunami caused by the Great East Japan Earthquake triggered severe accidents involving the units 1 to 4 at the Fukushima Dai-ichi nuclear power station (NPS). In order to re-operate existing nuclear power plants it should be necessary to reduce the core damage frequency on risk by tsunami. In this work, effects of the off-site power supply installation on resuming operation of nuclear power plants were investigated by utilizing the Tsunami-Level 1 Probability Risk Assessment (PRA). Unit 2 of the Onagawa nuclear power station, which resembled units 2 and 3 of Fukushima Dai-ichi, was selected for PRA. First, event-tree was created for the units of the Onagawa nuclear power station with the safety systems such as Emergency Core Cooling System (ECCS), investigating the plant situation at the time of the earthquake and tsunami occurrences. It was assumed that the magnitude of the tsunami was equivalent to the Great East Japan Earthquake. The accident-analytical progression-time was 36 hours, determined from the core-damage occurrence of the unit 3 of Fukushima Dai-ichi nuclear power station. Failure probabilities were calculated by the fault tree, which was created from the elements listed in the event tree. For the calculation, failure rates reported by the NUCIA (NUClear Information Archives) were primarily utilized. Then, obtained failure probabilities were embedded to the event tree. Core damage probabilities were evaluated by calculating success and failure rates for each accidental progression and scenarios. Restoration of the failed equipment and machineries was not considered in the analysis. Installation of the power supply vehicles at the nuclear power plant site reduced the core damage probability from 2.58×10 -6 to 8.56×10 -7 . However, continued addition of the power supply vehicles could not lower the core damage probability further more. In the case of Unit 2 of Onagawa nuclear power station, there could be a limit to lower the core damage

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.

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.

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)

ASME nuclear codes and standards risk management strategic plan

International Nuclear Information System (INIS)

Balkey, Kenneth R.

2003-01-01

Over the past 15 years, several risk-informed initiatives have been completed or are under development within the ASME Nuclear Codes and Standards organization. In order to better manage the numerous initiatives in the future, the ASME Board on Nuclear Codes and Standards has recently developed and approved a Risk Management Strategic Plan. This paper presents the latest approved version of the plan beginning with a background of applications completed to date, including the recent issuance of the ASME Standard for Probabilistic Risk Assessment (PRA) for Nuclear Power Plant Applications. The paper discusses potential applications within ASME Nuclear Codes and Standards that may require expansion of the PRA Standard, such as for new generation reactors, or the development of new PRA Standards. A long-term vision for the potential development and evolution to a nuclear systems code that adopts a risk-informed approach across a facility life-cycle (design, construction, operation, maintenance, and closure) is summarized. Finally, near term and long term actions are defined across the ASME Nuclear Codes and Standards organizations related to risk management, and related U.S. regulatory activities are also summarized. (author)

Use of probabilistic risk assessment in expert system usage for nuclear power plant safety

International Nuclear Information System (INIS)

Uhrig, R.E.

1987-01-01

The introduction of probability risk assessments (PRA's) to nuclear power plants in the Rasmussen Report (WASH-1400) gave us a means of evaluating the risk to the public associated with the operation of nuclear power plants, at least on a relative basis. While the choice of the ''source term'' and methodology in a PRA significantly influence the absolute probability and the consequences of core melt, comparison of two PRA calculations for two configurations of the same plant, carried out on a consistent basis, can be readily identify the increase in risk associated with going from one configuration of a plant to another by removing components or systems from service. This ratio of core melt probabilities (assuming no recovery of failed systems) obtained from two PRA calculations for different configurations was the criterion (called ''risk factor'') chosen as a basis for making a decision in an expert system as to what mitigating action, if any, would be taken to avoid a trip situation from developing. PRISIM was developed by JBF Associates of Knoxville under the sponsorship of the NRC as a system for Resident Inspectors at nuclear power plants to provide them with a relative safety status of the plant under all configurations. PRISIM calculated the risk factor---the ration of core melt probabilities of the plant under the current configuration relative to the normal configuration with all systems functioning---using an algorithm that emulates the results of the original PRA. It also presents time and core melt (assuming no recovery of systems or components)

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)

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)

Risk based modelling

International Nuclear Information System (INIS)

Chapman, O.J.V.; Baker, A.E.

1993-01-01

Risk based analysis is a tool becoming available to both engineers and managers to aid decision making concerning plant matters such as In-Service Inspection (ISI). In order to develop a risk based method, some form of Structural Reliability Risk Assessment (SRRA) needs to be performed to provide a probability of failure ranking for all sites around the plant. A Probabilistic Risk Assessment (PRA) can then be carried out to combine these possible events with the capability of plant safety systems and procedures, to establish the consequences of failure for the sites. In this way the probability of failures are converted into a risk based ranking which can be used to assist the process of deciding which sites should be included in an ISI programme. This paper reviews the technique and typical results of a risk based ranking assessment carried out for nuclear power plant pipework. (author)

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

An evaluation of risk methods for prioritizing fire protection features: a procedure for fire barrier penetration seals

International Nuclear Information System (INIS)

Dey, M.K.

2004-01-01

This paper generally evaluates risk methods available for prioritizing fire protection features. Risk methods involving both the use of qualitative insights, and quantitative results from a fire probabilistic risk analysis are reviewed. The applicability of these methods to develop a prioritized list of fire barrier penetration seals in a plant based on risk significance is presented as a procedure to illustrate the benefits of the methods. The paper concludes that current fire risk assessment methods can be confidently used to prioritize plant fire protection features, specifically fire barrier penetration seals. Simple prioritization schemes, using qualitative assessments and insights from fire PRA methodology may be implemented without the need for quantitative results. More elaborate prioritization schemes that allow further refinements to the categorization process may be implemented using the quantitative results of the screening processes in good fire PRAs. The use of the quantitative results from good fire PRAs provide several benefits for risk prioritization of fire protection features at plants, mainly from the plant systems analyses conducted for a fire PRA

Analysis of climate and anthropogenic impacts on runoff in the Lower Pra River Basin of Ghana.

Science.gov (United States)

Awotwi, Alfred; Anornu, Geophrey Kwame; Quaye-Ballard, Jonathan; Annor, Thompson; Forkuo, Eric Kwabena

2017-12-01

The Lower Pra River Basin (LPRB), located in the forest zone of southern Ghana has experienced changes due to variability in precipitation and diverse anthropogenic activities. Therefore, to maintain the functions of the ecosystem for water resources management, planning and sustainable development, it is important to differentiate the impacts of precipitation variability and anthropogenic activities on stream flow changes. We investigated the variability in runoff and quantified the contributions of precipitation and anthropogenic activities on runoff at the LPRB. Analysis of the precipitation-runoff for the period 1970-2010 revealed breakpoints in 1986, 2000, 2004 and 2010 in the LPRB. The periods influenced by anthropogenic activities were categorized into three periods 1987-2000, 2001-2004 and 2005-2010, revealing a decrease in runoff during 1987-2000 and an increase in runoff during 2001-2004 and 2005-2010. Assessment of monthly, seasonal and annual runoff depicted a significant increasing trend in the runoff time series during the dry season. Generally, runoff increased at a rate of 9.98 × 10 7 m 3 yr -1 , with precipitation variability and human activities contributing 17.4% and 82.3% respectively. The dominant small scale alluvial gold mining activity significantly contributes to the net runoff variability in LPRB.

Perspective on US NRC Policy Issues Concerning Use of Risk Insights for Non-LWR

International Nuclear Information System (INIS)

Ha, Jun Su; Kim, In Goo; Huh, Chang Wook; Kim, Kyun Tae

2011-01-01

Since the PRA Implementation plan of US NRC (1994), PRA has been applied to all NPPs in USA and risk insights have been used for the regulation as a complement of the deterministic approaches. RIRIP (Risk-Informed Regulation Implementation Plan, 2000) and RPP (Risk-Informed and Performance-Based Plan, 2007) were announced by US NRC thereafter, which recommended enhanced use of risk insights. In the meantime, there have been lots of policy issues concerning use of risk insights for licensing Non-LWR designs, which will be discussed in this paper to understand the stream of perspectives on US NRC's approach

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

A Research Roadmap for Computation-Based Human Reliability Analysis

Energy Technology Data Exchange (ETDEWEB)

Boring, Ronald [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Joe, Jeffrey [Idaho National Lab. (INL), Idaho Falls, ID (United States); Smith, Curtis [Idaho National Lab. (INL), Idaho Falls, ID (United States); Groth, Katrina [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States)

2015-08-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

A Research Roadmap for Computation-Based Human Reliability Analysis

International Nuclear Information System (INIS)

Boring, Ronald; Mandelli, Diego; Joe, Jeffrey; Smith, Curtis; Groth, Katrina

2015-01-01

The United States (U.S.) Department of Energy (DOE) is sponsoring research through the Light Water Reactor Sustainability (LWRS) program to extend the life of the currently operating fleet of commercial nuclear power plants. The Risk Informed Safety Margin Characterization (RISMC) research pathway within LWRS looks at ways to maintain and improve the safety margins of these plants. The RISMC pathway includes significant developments in the area of thermalhydraulics code modeling and the development of tools to facilitate dynamic probabilistic risk assessment (PRA). PRA is primarily concerned with the risk of hardware systems at the plant; yet, hardware reliability is often secondary in overall risk significance to human errors that can trigger or compound undesirable events at the plant. This report highlights ongoing efforts to develop a computation-based approach to human reliability analysis (HRA). This computation-based approach differs from existing static and dynamic HRA approaches in that it: (i) interfaces with a dynamic computation engine that includes a full scope plant model, and (ii) interfaces with a PRA software toolset. The computation-based HRA approach presented in this report is called the Human Unimodels for Nuclear Technology to Enhance Reliability (HUNTER) and incorporates in a hybrid fashion elements of existing HRA methods to interface with new computational tools developed under the RISMC pathway. The goal of this research effort is to model human performance more accurately than existing approaches, thereby minimizing modeling uncertainty found in current plant risk models.

Research needs for risk-informed, performance-based regulation

International Nuclear Information System (INIS)

Bailey, J.A.

1997-01-01

Palo Verde Nuclear Generating Station has used PRA-derived risk insights for about 10 years now. The plant originally started applying PRA modeling to an auxiliary feedwater system during the initial licensing phases of the plant, and as a result of that, they were able to work with the NRC and apply some graded quality requirements to that particular system. There was a third redundant auxiliary feedwater pump, and they now can treat that system as partially safety related and partially non-safety related. So it was an advance for Palo Verde at that time to be able to make decisions with a PRA and they began learning how to use those techniques. After completing the IPE it became natural for the plant to make a transition into other areas at the plant to look for areas where the insights gained from PRA could be applied into their decision-making processes. Those that the plant embarked upon initially were areas where they could gain operational risk assessment insights. The author goes on to discuss experiences gained in using these techniques to better assess the safety of operations within the plant. In addition he offers comments on areas which need further development and research to make them more applicable to a plant by plant basis

Treatment of system dependencies and human interactions in PRA studies: a review and sensitivity study

International Nuclear Information System (INIS)

Orvis, D.D.; Joksimovich, V.; Worledge, D.H.

1985-01-01

The Electric Power Research Institute sponsored the review and comparison of five PRA studies: Arkansas Nuclear One - Unit 1, Big Rock Point, Grand Gulf, Limerick, and Zion - Unit 1. The review has been conducted in two phases. The Phase I review may be characterized as a qualitative look into many aspects of a PRA study. The Phase II review was performed to quantify the extent that differences in analytical techniques or key assumptions in these areas affect the differences in study results. In each of the PRA studies reviewed, the general descriptions of analytical approaches and descriptions of the analyses of event tree, fault tree and human interaction analyses that affected the dominant core damage sequences were reviewed. When these descriptions aroused interest because of seeming inconsistencies within the study or with other studies, they were pursued in some depth. The approaches or assumptions were contrasted to similar elements from other studies, and sensitivity analyses were performed in many cases to test the significance of results to the analytical models or assumptions. Inferences were drawn from the results regarding significance of the item to plant-specific results and, where possible, were generalized to other PRAs. This paper describes the results of the review of system dependencies and human interactions

Risk-based technical specifications: Development and application of an approach to the generation of a plant specific real-time risk model

International Nuclear Information System (INIS)

Puglia, B.; Gallagher, D.; Amico, P.; Atefi, B.

1992-10-01

This report describes a process developed to convert an existing PRA into a model amenable to real time, risk-based technical specification calculations. In earlier studies (culminating in NUREG/CR-5742), several risk-based approaches to technical specification were evaluated. A real-time approach using a plant specific PRA capable of modeling plant configurations as they change was identified as the most comprehensive approach to control plant risk. A master fault tree logic model representative of-all of the core damage sequences was developed. Portions of the system fault trees were modularized and supercomponents comprised of component failures with similar effects were developed to reduce the size of the model and, quantification times. Modifications to the master fault tree logic were made to properly model the effect of maintenance and recovery actions. Fault trees representing several actuation systems not modeled in detail in the existing PRA were added to the master fault tree logic. This process was applied to the Surry NUREG-1150 Level 1 PRA. The master logic mode was confirmed. The model was then used to evaluate frequency associated with several plant configurations using the IRRAS code. For all cases analyzed computational time was less than three minutes. This document Volume 2, contains appendices A, B, and C. These provide, respectively: Surry Technical Specifications Model Database, Surry Technical Specifications Model, and a list of supercomponents used in the Surry Technical Specifications Model

Controlling principles for prior probability assignments in nuclear risk assessment

International Nuclear Information System (INIS)

Cook, I.; Unwin, S.D.

1986-01-01

As performed conventionally, nuclear probabilistic risk assessment (PRA) may be criticized as utilizing inscrutable and unjustifiably ''precise'' quantitative informed judgment or extrapolation from that judgment. To meet this criticism, controlling principles that govern the formulation of probability densities are proposed, given only the informed input that would be required for a simple bounding analysis. These principles are founded upon information theoretic ideas of maximum uncertainty and cover both cases in which there exists a stochastic model of the phenomenon of interest and cases in which these is no such model. In part, the principles are conventional, and such an approach is justified by appealing to certain analogies in accounting practice and judicial decision making. Examples are given. Appropriate employment of these principles is expected to facilitate substantial progress toward PRA scrutability and transparency

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.

The Importance of HRA in Human Space Flight: Understanding the Risks

Science.gov (United States)

Hamlin, Teri

2010-01-01

Human performance is critical to crew safety during space missions. Humans interact with hardware and software during ground processing, normal flight, and in response to events. Human interactions with hardware and software can cause Loss of Crew and/or Vehicle (LOCV) through improper actions, or may prevent LOCV through recovery and control actions. Humans have the ability to deal with complex situations and system interactions beyond the capability of machines. Human Reliability Analysis (HRA) is a method used to qualitatively and quantitatively assess the occurrence of human failures that affect availability and reliability of complex systems. Modeling human actions with their corresponding failure probabilities in a Probabilistic Risk Assessment (PRA) provides a more complete picture of system risks and risk contributions. A high-quality HRA can provide valuable information on potential areas for improvement, including training, procedures, human interfaces design, and the need for automation. Modeling human error has always been a challenge in part because performance data is not always readily available. For spaceflight, the challenge is amplified not only because of the small number of participants and limited amount of performance data available, but also due to the lack of definition of the unique factors influencing human performance in space. These factors, called performance shaping factors in HRA terminology, are used in HRA techniques to modify basic human error probabilities in order to capture the context of an analyzed task. Many of the human error modeling techniques were developed within the context of nuclear power plants and therefore the methodologies do not address spaceflight factors such as the effects of microgravity and longer duration missions. This presentation will describe the types of human error risks which have shown up as risk drivers in the Shuttle PRA which may be applicable to commercial space flight. As with other large PRAs

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)

A technique for human error analysis (ATHEANA)

International Nuclear Information System (INIS)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W.

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions

A technique for human error analysis (ATHEANA)

Energy Technology Data Exchange (ETDEWEB)

Cooper, S.E.; Ramey-Smith, A.M.; Wreathall, J.; Parry, G.W. [and others

1996-05-01

Probabilistic risk assessment (PRA) has become an important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. Human reliability analysis (HRA) is a critical element of PRA; however, limitations in the analysis of human actions in PRAs have long been recognized as a constraint when using PRA. A multidisciplinary HRA framework has been developed with the objective of providing a structured approach for analyzing operating experience and understanding nuclear plant safety, human error, and the underlying factors that affect them. The concepts of the framework have matured into a rudimentary working HRA method. A trial application of the method has demonstrated that it is possible to identify potentially significant human failure events from actual operating experience which are not generally included in current PRAs, as well as to identify associated performance shaping factors and plant conditions that have an observable impact on the frequency of core damage. A general process was developed, albeit in preliminary form, that addresses the iterative steps of defining human failure events and estimating their probabilities using search schemes. Additionally, a knowledge- base was developed which describes the links between performance shaping factors and resulting unsafe actions.

Application of probabilistic risk assessment: Evaluating remedial alternatives at the Portland Harbor Superfund Site, Portland, Oregon, USA.

Science.gov (United States)

Ruffle, Betsy; Henderson, James; Murphy-Hagan, Clare; Kirkwood, Gemma; Wolf, Frederick; Edwards, Deborah A

2018-01-01

A probabilistic risk assessment (PRA) was performed to evaluate the range of potential baseline and postremedy health risks to fish consumers at the Portland Harbor Superfund Site (the "Site"). The analysis focused on risks of consuming fish resident to the Site containing polychlorinated biphenyls (PCBs), given that this exposure scenario and contaminant are the primary basis for US Environmental Protection Agency's (USEPA's) selected remedy per the January 2017 Record of Decision (ROD). The PRA used probability distributions fit to the same data sets used in the deterministic baseline human health risk assessment (BHHRA) as well as recent sediment and fish tissue data to evaluate the range and likelihood of current baseline cancer risks and noncancer hazards for anglers. Areas of elevated PCBs in sediment were identified on the basis of a geospatial evaluation of the surface sediment data, and the ranges of risks and hazards associated with pre- and postremedy conditions were calculated. The analysis showed that less active remediation (targeted to areas with the highest concentrations) compared to the remedial alternative selected by USEPA in the ROD can achieve USEPA's interim risk management benchmarks (cancer risk of 10 -4 and noncancer hazard index [HI] of 10) immediately postremediation for the vast majority of subsistence anglers that consume smallmouth bass (SMB) fillet tissue. In addition, the same targeted remedy achieves USEPA's long-term benchmarks (10 -5 and HI of 1) for the majority of recreational anglers. Additional sediment remediation would result in negligible additional risk reduction due to the influence of background. The PRA approach applied here provides a simple but adaptive framework for analysis of risks and remedial options focused on variability in exposures. It can be updated and refined with new data to evaluate and reduce uncertainty, improve understanding of the Site and target populations, and foster informed remedial decision

Optimizing risk management

International Nuclear Information System (INIS)

Kindred, G.W.

2000-01-01

Commercial nuclear power plant management is focussed on the safe, efficient, economical production of electricity. To accomplish the safe aspect of the equation, risk must be determined for the operation and maintenance of the facility. To accomplish the efficient aspect of the equation, management must understand those risks and factor risk insights into their decision process. The final piece of the equation is economical which is accomplished by minimizing, plant outage durations and proper utilization of resources. Probabilistic Risk Assessment can provide the risk insights to accomplish all three; safety, efficiency, and economically. How? Safe production of electricity can be quantified by use of PRA modeling and other risk insights that can determine the core damage frequency. Efficient production of electricity can be influenced by providing management with quantified risk insights for use in decision making. And, one example of economical production of electricity is by not having over conservative deterministic based defense in depth approaches to system maintenance and availability. By using risk-informed insights nuclear safety can be quantified and risk can be managed. Confidence in this approach can be achieved by ensuring the content and quality of the PRA is standardized throughout the industry. The time has arrived for Probabilistic Risk Assessment to take an active position as a major role player in the safe, efficient, and economical operation of commercial nuclear power plants. (author)

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE)

International Nuclear Information System (INIS)

C. L. Smith

2006-01-01

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 is primarily 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. SAPHIRE can be used to model a complex system's response to initiating events and quantify associated consequential outcome frequencies. Specifically, for nuclear power plant applications, SAPHIRE can identify important contributors to core damage (Level 1 PRA) and containment failure during a severe accident which lead to releases (Level 2 PRA). It can be used for a PRA where the reactor is at full power, low power, or at shutdown conditions. Furthermore, it can be used to analyze both internal and external initiating events and has special features for transforming an internal events model to a model for external events, such as flooding and fire analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to the public and environment (Level 3 PRA). SAPHIRE also includes a separate module called the Graphical Evaluation Module (GEM). GEM is a special user interface linked to SAPHIRE that automates the SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events (for example, to calculate a conditional core damage probability) very efficiently and expeditiously. This report provides an overview of the functions

Development of regulatory guidance for risk-informing digital system reviews

International Nuclear Information System (INIS)

Arndt, S. A.

2006-01-01

In 1995, the U.S. Nuclear Regulatory Commission (NRC) issued the Probabilistic Risk Assessment (PRA) Policy Statement, which encourages the increased use of PRA and associated analyses in all regulatory matters to the extent supported by the state-of-the-art in PRA and the data. This policy applies, in part, to the review of digital systems, which offer the potential to improve plant safety and reliability through such features as increased hardware reliability and stability and improved failure detection capability. However, there are presently no universally accepted methods for modeling digital systems in current-generation PRAs. Further, there are ongoing debates among the PRA technical community regarding the level of detail that any digital system reliability model must have to adequately model the complex system interactions that can contribute to digital system failure modes. Moreover, for PRA modeling of digital reactor protection and control systems, direct interactions between system components and indirect interactions through controlled/supervised plant processes may necessitate the use of dynamic PRA methodologies. This situation has led the NRC to consider developing performance based rather than prescriptive regulatory guidance in this area. This paper will discuss the development of this guidance and some preliminary concepts. (authors)

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).

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

Guideliness for system modeling: fault tree [analysis

Energy Technology Data Exchange (ETDEWEB)

Lee, Yoon Hwan; Yang, Joon Eon; Kang, Dae Il; Hwang, Mee Jeong

2004-07-01

This document, the guidelines for system modeling related to Fault Tree Analysis(FTA), is intended to provide the guidelines with the analyzer to construct the fault trees in the level of the capability category II of ASME PRA standard. Especially, they are to provide the essential and basic guidelines and the related contents to be used in support of revising the Ulchin 3 and 4 PSA model for risk monitor within the capability category II of ASME PRA standard. Normally the main objective of system analysis is to assess the reliability of system modeled by Event Tree Analysis (ETA). A variety of analytical techniques can be used for the system analysis, however, FTA method is used in this procedures guide. FTA is the method used for representing the failure logic of plant systems deductively using AND, OR or NOT gates. The fault tree should reflect all possible failure modes that may contribute to the system unavailability. This should include contributions due to the mechanical failures of the components, Common Cause Failures (CCFs), human errors and outages for testing and maintenance. This document identifies and describes the definitions and the general procedures of FTA and the essential and basic guidelines for reving the fault trees. Accordingly, the guidelines for FTA will be capable to guide the FTA to the level of the capability category II of ASME PRA standard.

Guideliness for system modeling: fault tree [analysis

International Nuclear Information System (INIS)

Lee, Yoon Hwan; Yang, Joon Eon; Kang, Dae Il; Hwang, Mee Jeong

2004-07-01

This document, the guidelines for system modeling related to Fault Tree Analysis(FTA), is intended to provide the guidelines with the analyzer to construct the fault trees in the level of the capability category II of ASME PRA standard. Especially, they are to provide the essential and basic guidelines and the related contents to be used in support of revising the Ulchin 3 and 4 PSA model for risk monitor within the capability category II of ASME PRA standard. Normally the main objective of system analysis is to assess the reliability of system modeled by Event Tree Analysis (ETA). A variety of analytical techniques can be used for the system analysis, however, FTA method is used in this procedures guide. FTA is the method used for representing the failure logic of plant systems deductively using AND, OR or NOT gates. The fault tree should reflect all possible failure modes that may contribute to the system unavailability. This should include contributions due to the mechanical failures of the components, Common Cause Failures (CCFs), human errors and outages for testing and maintenance. This document identifies and describes the definitions and the general procedures of FTA and the essential and basic guidelines for reving the fault trees. Accordingly, the guidelines for FTA will be capable to guide the FTA to the level of the capability category II of ASME PRA standard

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

"VEM PRA RUA": THE POLITICAL AND THE POLITICS ON THE WEB

Directory of Open Access Journals (Sweden)

Benedito Fernando Pereira

2014-12-01

Full Text Available Considering the political and the social division of senses, this paper seeks to check how is the process of signification in a society increasingly challenged by electronic discourse and the ways in which political and policy find themselves affected by it. In order do that, we make the analysis of discursive statements “vem pra rua” and “somos a rede social” that were present in banners and posters in street protests in Brazil in 2013, which were organized and delivered, in large part, by virtual means. We observed that such utterances go through a process of appropriation and re-signification, with shifting meanings that run of market logic, go to the social politics and go back to the market logic. We had also observed how the urban environment is affected by the electronic discourse that now constitutes the ways it makes sense.

A Quantitative Accident Sequence Analysis for a VHTR

Energy Technology Data Exchange (ETDEWEB)

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

2016-05-15

In Korea, the basic design features of VHTR are currently discussed in the various design concepts. Probabilistic risk assessment (PRA) offers a logical and structured method to assess risks of a large and complex engineered system, such as a nuclear power plant. It will be introduced at an early stage in the design, and will be upgraded at various design and licensing stages as the design matures and the design details are defined. Risk insights to be developed from the PRA are viewed as essential to developing a design that is optimized in meeting safety objectives and in interpreting the applicability of the existing demands to the safety design approach of the VHTR. In this study, initiating events which may occur in VHTRs were selected through MLD method. The initiating events were then grouped into four categories for the accident sequence analysis. Initiating events frequency and safety systems failure rate were calculated by using reliability data obtained from the available sources and fault tree analysis. After quantification, uncertainty analysis was conducted. The SR and LR frequency are calculated respectively 7.52E- 10/RY and 7.91E-16/RY, which are relatively less than the core damage frequency of LWRs.

An Initiating-Event Analysis for PSA of Hanul Units 3 and 4: Results and Insights

International Nuclear Information System (INIS)

Kim, Dong-San; Park, Jin Hee

2015-01-01

As a part of the PSA, an initiating-event (IE) analysis was newly performed by considering the current state of knowledge and the requirements of the ASME/ANS probabilistic risk assessment (PRA) standard related to IE analysis. This paper describes the methods of, results and some insights from the IE analysis for the PSA of the Hanul units 3 and 4. In this study, as a part of the PSA for the Hanul units 3 and 4, an initiating-event (IE) analysis was newly performed by considering the current state of knowledge and the requirements of the ASME/ANS probabilistic risk assessment (PRA) standard. In comparison with the previous IE analysis, this study performed a more systematic and detailed analysis to identify potential initiating events, and calculated the IE frequencies by using the state-of-the-art methods and the latest data. As a result, not a few IE frequencies are quite different from the previous frequencies, which can change the major accident sequences obtained from the quantification of the PSA model

Activity risk coefficients for living generations

International Nuclear Information System (INIS)

Raicevic, J.; Merkle, M.; Ninkovic, M. M.

1993-01-01

This paper deals with the new concept of the Activity risk coefficients, ARCs, which are in Probabilistic risk assessment PRA computer codes used for the calculation of the stochastic effects due to low dose exposures. As an example, ARC expressions for the Cloudshine is derived. (author)

Track 6: safety and risk management. Plant operational risk management. Plant Configuration Risk Assessment Methodology Development for Periodic Maintenance

International Nuclear Information System (INIS)

Yang, Huichang; Chung, Chang Hyun; Sung, Key Yong

2001-01-01

plant risk level. Such a change in the arrangement of the plant equipment and system at a given time period can be represented as the plant configuration. The plant configuration risk assessment methodology that was developed during this study consists of six steps, as follows: 1. Identification of plant configuration: In this step, various events that occurred in the plant should be identified through a review of the plant operation records such as the periodic maintenance and inspection schedules, maintenance or repair request logs, trouble reports, and other documents related to operational activity. 2. Evaluation of probabilistic risk assessment (PRA) model and computer codes: For the effective evaluation of plant risk during normal operation, an appropriate plant risk model should 273 be used, and the capability of computer codes should be evaluated. There might be numerous events that require the maintenance activity during normal operation. To handle these events during the risk calculation, an optimized plant PRA model and a risk analysis tool of fast calculation capacity are needed. 3. Development of baseline risk model and evaluation of baseline risk: The baseline risk model is a risk model similar to that used for the level 1 PRA, but the maintenance-related events are excluded. This methodology focuses on the relative risk change caused by the usual plant events. For this purpose, the baseline risk that will be the reference of risk variation should be evaluated as reasonably as possible. 4. Analysis of components and systems: For the detailed risk analysis, it is useful to perform the importance analysis for the target components or systems before calculating the plant risks. In terms of system unavailability analysis and importance analysis, information of specific components and systems should be performed for the detailed risk analysis 5. Evaluation of configuration risks and sensitivity analysis: Using the configuration and the system information from

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

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

International Nuclear Information System (INIS)

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

1990-06-01

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

Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE), Version 5.0: Integrated Reliability and Risk Analysis System (IRRAS) reference manual. Volume 2

International Nuclear Information System (INIS)

Russell, K.D.; Kvarfordt, K.J.; Skinner, N.L.; Wood, S.T.; Rasmuson, D.M.

1994-07-01

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) refers to a set of several microcomputer programs that were developed to create and analyze probabilistic risk assessments (PRAs), primarily for nuclear power plants. The Integrated Reliability and Risk Analysis System (IRRAS) is a state-of-the-art, microcomputer-based probabilistic risk assessment (PRA) model development and analysis tool to address key nuclear plant safety issues. IRRAS is an integrated software tool that gives the use the ability to create and analyze fault trees and accident sequences using a microcomputer. This program provides functions that range from graphical fault tree construction to cut set generation and quantification to report generation. Version 1.0 of the IRRAS program was released in February of 1987. Since then, many user comments and enhancements have been incorporated into the program providing a much more powerful and user-friendly system. This version has been designated IRRAS 5.0 and is the subject of this Reference Manual. Version 5.0 of IRRAS provides the same capabilities as earlier versions and ads the ability to perform location transformations, seismic analysis, and provides enhancements to the user interface as well as improved algorithm performance. Additionally, version 5.0 contains new alphanumeric fault tree and event used for event tree rules, recovery rules, and end state partitioning

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

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

ATHEANA: open-quotes a technique for human error analysisclose quotes entering the implementation phase

International Nuclear Information System (INIS)

Taylor, J.; O'Hara, J.; Luckas, W.

1997-01-01

Probabilistic Risk Assessment (PRA) has become an increasingly important tool in the nuclear power industry, both for the Nuclear Regulatory Commission (NRC) and the operating utilities. The NRC recently published a final policy statement, SECY-95-126, encouraging the use of PRA in regulatory activities. Human reliability analysis (HRA), while a critical element of PRA, has limitations in the analysis of human actions in PRAs that have long been recognized as a constraint when using PRA. In fact, better integration of HRA into the PRA process has long been a NRC issue. Of particular concern, has been the omission of errors of commission - those errors that are associated with inappropriate interventions by operators with operating systems. To address these concerns, the NRC identified the need to develop an improved HRA method, so that human reliability can be better represented and integrated into PRA modeling and quantification. The purpose of the Brookhaven National Laboratory (BNL) project, entitled 'Improved HRA Method Based on Operating Experience' is to develop a new method for HRA which is supported by the analysis of risk-significant operating experience. This approach will allow a more realistic assessment and representation of the human contribution to plant risk, and thereby increase the utility of PRA. The project's completed, ongoing, and future efforts fall into four phases: (1) Assessment phase (FY 92/93); (2) Analysis and Characterization phase (FY 93/94); (3) Development phase (FY 95/96); and (4) Implementation phase (FY 96/97 ongoing)

Auxiliary feedwater system risk-based inspection guide for the North Anna nuclear power plants

International Nuclear Information System (INIS)

Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

1992-10-01

In a study sponsored by the US Nuclear regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. North Anna was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the North Anna plant

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

International Nuclear Information System (INIS)

Vierow, Karen; Aldemir, Tunc

2009-01-01

The project entitled, 'Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors', was conducted as a DOE NERI project collaboration between Texas A and M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors: Final Scientific/Technical Report

Energy Technology Data Exchange (ETDEWEB)

Vierow, Karen; Aldemir, Tunc

2009-09-10

The project entitled, “Uncertainty Quantification in the Reliability and Risk Assessment of Generation IV Reactors”, was conducted as a DOE NERI project collaboration between Texas A&M University and The Ohio State University between March 2006 and June 2009. The overall goal of the proposed project was to develop practical approaches and tools by which dynamic reliability and risk assessment techniques can be used to augment the uncertainty quantification process in probabilistic risk assessment (PRA) methods and PRA applications for Generation IV reactors. This report is the Final Scientific/Technical Report summarizing the project.

Sequence variations and protein expression levels of the two immune evasion proteins Gpm1 and Pra1 influence virulence of clinical Candida albicans isolates.

Science.gov (United States)

Luo, Shanshan; Hipler, Uta-Christina; Münzberg, Christin; Skerka, Christine; Zipfel, Peter F

2015-01-01

Candida albicans, the important human fungal pathogen uses multiple evasion strategies to control, modulate and inhibit host complement and innate immune attack. Clinical C. albicans strains vary in pathogenicity and in serum resistance, in this work we analyzed sequence polymorphisms and variations in the expression levels of two central fungal complement evasion proteins, Gpm1 (phosphoglycerate mutase 1) and Pra1 (pH-regulated antigen 1) in thirteen clinical C. albicans isolates. Four nucleotide (nt) exchanges, all representing synonymous exchanges, were identified within the 747-nt long GPM1 gene. For the 900-nt long PRA1 gene, sixteen nucleotide exchanges were identified, which represented synonymous, as well as non-synonymous exchanges. All thirteen clinical isolates had a homozygous exchange (A to G) at position 73 of the PRA1 gene. Surface levels of Gpm1 varied by 8.2, and Pra1 levels by 3.3 fold in thirteen tested isolates and these differences influenced fungal immune fitness. The high Gpm1/Pra1 expressing candida strains bound the three human immune regulators more efficiently, than the low expression strains. The difference was 44% for Factor H binding, 51% for C4BP binding and 23% for plasminogen binding. This higher Gpm1/Pra1 expressing strains result in enhanced survival upon challenge with complement active, Factor H depleted human serum (difference 40%). In addition adhesion to and infection of human endothelial cells was increased (difference 60%), and C3b surface deposition was less effective (difference 27%). Thus, variable expression levels of central immune evasion protein influences immune fitness of the human fungal pathogen C. albicans and thus contribute to fungal virulence.

Risk, probability and uncertainty in the calculations of gas cooled reactor of PBMR type. Part 2

International Nuclear Information System (INIS)

Serbanescu, Dan

2004-01-01

The paper presents the main conclusions of the insights to a cooled gas reactor from the perspective of the following notions: probability, uncertainty, entropy and risk. Some results of the on-going comparison between the insights obtained from three models and approaches are presented. The approaches consider the Pebble Bed Module Reactor (PBMR) NPP as a thermodynamic installation and as hierarchical system with or without considering the information exchange between its various levels. The existing model was a basis for a PRA going on in phases for PBMR. In the first part of this paper results from phase II of this PRA were presented. Further activities going on in the preparation for phase II PRA and for the development of a specific application of using PRA during the design phases for PBMR are undergoing with some preliminary results and conclusions. However, for the purposes of this paper and the comparative review of various models in the part two one presents the risk model (model B) based on the assumption and ideas laid down at the basis of the future inter-comparison of this model with other plant models. The assumptions concern: the uncertainties for the quantification of frequencies; list of initiated events; interfaces with the deterministic calculation; integrated evaluation of all the plant states; risk of the release of radionuclide; the balance between the number and function of the active systems and the passive systems; systems interdependencies in PBMR PRA; use of PRA for the evaluation of the impact of various design changes on plant risk. The model B allows basically evaluating the level of risk of the plant by calculating it as a result of acceptance challenge to the plant. By using this model the departure from a reference state is given by the variation in the risk metrics adopted for the study. The paper present also the synergetic model (model C). The evaluation of risk in the model C is considering also the information process. The

Pulsa o coração da cidade: errâncias, afectos e potências no dia e na noite da Praça do Ferreira

Directory of Open Access Journals (Sweden)

Alice Dote

2017-12-01

Full Text Available O presente artigo aborda a potência dos usos, contra-usos e modos de habitar dos artistas de rua da Praça do Ferreira, na cidade de Fortaleza, Ceará. O trabalho apoia-se nos percursos e nas errâncias urbanas da vivência na e da Praça do Ferreira em diferentes temporalidades (diurna e noturna, especialmente no contexto de apresentações noturnas do Grupo As 10 Graças de Palhaçaria aos moradores da Praça. Através desses que têm a rua como casa, agem pelas brechas e proliferam-se pelas margens, proponho-me a perceber a potência da arte de rua, do encontro e da experiência de alteridade na Praça do Ferreira. Finalizo o texto apontando que esse local, assim ocupado, se impregna de significados outros e revela-se como um território de criação, de inventividade, de existência e resistência, portanto, de potência de vida que é, em si, potência política. Palavras-chave: Praça do Ferreira; Fortaleza; cidade; arte urbana; artista de rua

Using risk based tools in emergency response

International Nuclear Information System (INIS)

Dixon, B.W.; Ferns, K.G.

1987-01-01

Probabilistic Risk Assessment (PRA) techniques are used by the nuclear industry to model the potential response of a reactor subjected to unusual conditions. The knowledge contained in these models can aid in emergency response decision making. This paper presents requirements for a PRA based emergency response support system to date. A brief discussion of published work provides background for a detailed description of recent developments. A rapid deep assessment capability for specific portions of full plant models is presented. The program uses a screening rule base to control search space expansion in a combinational algorithm

Guidelines on the scope, content, and use of comprehensive risk assessment in the management of high-level nuclear waste transportation

International Nuclear Information System (INIS)

Golding, D.; White, A.

1990-12-01

This report discusses the scope of risk assessment strategies in the management of the transport of high-level radioactive wastes. In spite of the shortcomings of probabilistic risk assessment(PRA), the Transportation Needs Assessment recommended this as the preferred methodology to assess the risks of high level nuclear waste (HLNW) transportation. A PRA also will need to heed the lessons learned from the development and application of PRA elsewhere, such as in the nuclear power industry. A set of guidelines will aid this endeavor by outlining the appropriate scope, content, and use of a risk assessment which is more responsive to the uncertainties, human-technical interactions, social forces, and iterative relationship with risk management strategies, than traditional PRAS. This more expansive definition, which encompasses but is not totally reliant on rigorous data requirements and quantitative probability estimates, we term Comprehensive Risk Assessment (CRA) Guidelines will be developed in three areas: the limitations of existing methodologies and suggested modifications; CRA as part of a flexible, effective, adaptive risk management system for HLNW transportation; and, the use of CRA in risk communication

The application of availability analysis to nuclear power plants

International Nuclear Information System (INIS)

Brooks, A.C.

1984-01-01

The use of probabilistic risk analysis (PRA) to assess the risks from nuclear power plants is now well established. Considerably less attention has been given so far to the use of availability analysis techniques. The economics of power generation are now such that with nuclear power currently supplying a substantial fraction of power in many countries, increasing attention is being paid to improving plant availability. This paper presents a technique for systematically identifying the areas in which measures to improve plant availability will be most effective. (author)

Effect of antigravity suit inflation on cardiovascular, PRA, and PVP responses in humans.

Science.gov (United States)

Kravik, S E; Keil, L C; Geelen, G; Wade, C E; Barnes, P R; Spaul, W A; Elder, C A; Greenleaf, J E

1986-08-01

Blood pressure, pulse rate (PR), serum osmolality and electrolytes, as well as plasma vasopressin (PVP) and plasma renin activity (PRA), were measured in five men and two women [mean age 38.6 +/- 3.9 (SE) yr] before, during, and after inflation of an antigravity suit that covered the legs and abdomen. After 24 h of fluid deprivation the subjects stood quietly for 3 h: the 1st h without inflation, the 2nd with inflation to 60 Torr, and the 3rd without inflation. A similar control noninflation experiment was conducted 10 mo after the inflation experiment using five of the seven subjects except that the suit was not inflated during the 3-h period. Mean arterial pressure increased by 14 +/- 4 (SE) Torr (P less than 0.05) with inflation and decreased by 15 +/- 5 Torr (P less than 0.05) after deflation. Pulse pressure (PP) increased by 7 +/- 2 Torr (P less than 0.05) with inflation and PR decreased by 11 +/- 5 beats/min (P less than 0.05); PP and PR returned to preinflation levels after deflation. Plasma volume decreased by 6.1 +/- 1.5% and 5.3 +/- 1.6% (P less than 0.05) during hours 1 and 3, respectively, and returned to base line during inflation. Inflation decreased PVP from 6.8 +/- 1.1 to 5.6 +/- 1.4 pg/ml (P less than 0.05) and abolished the significant rise in PRA during hour 1. Both PVP and PRA increased significantly after deflation: delta = 18.0 +/- 5.1 pg/ml and 4.34 +/- 1.71 ng angiotensin I X ml-1 X h-1, respectively. Serum osmolality and Na+ and K+ concentrations were unchanged during the 3 h of standing.(ABSTRACT TRUNCATED AT 250 WORDS)

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

IEEE guide for the analysis of human reliability

International Nuclear Information System (INIS)

Dougherty, E.M. Jr.

1987-01-01

The Institute of Electrical and Electronics Engineers (IEEE) working group 7.4 of the Human Factors and Control Facilities Subcommittee of the Nuclear Power Engineering Committee (NPEC) has released its fifth draft of a Guide for General Principles of Human Action Reliability Analysis for Nuclear Power Generating Stations, for approval of NPEC. A guide is the least mandating in the IEEE hierarchy of standards. The purpose is to enhance the performance of an human reliability analysis (HRA) as a part of a probabilistic risk assessment (PRA), to assure reproducible results, and to standardize documentation. The guide does not recommend or even discuss specific techniques, which are too rapidly evolving today. Considerable maturation in the analysis of human reliability in a PRA context has taken place in recent years. The IEEE guide on this subject is an initial step toward bringing HRA out of the research and development arena into the toolbox of standard engineering practices

Reviewing the development of an artificial intelligence based risk program

International Nuclear Information System (INIS)

Dixon, B.W.; Hinton, M.F.

1985-01-01

A successful application of nonconventional programming methods has been achieved in computer-assisted probabilistic risk assessment (PRA). The event tree sequence importance calculator, SQUIMP, provides for prompted data entry, generic expansion, on-line pruning, boolean reductions, and importance factor selection. SQUIMP employs constructs typically found in artificial intelligence (AI) programs. The development history of SQUIMP is outlined and its internal structure described as background for a discussion on the applicability of symbolic programming methods in PRA

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

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

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

Auxiliary feedwater system risk-based inspection guide for the Ginna Nuclear Power Plant

International Nuclear Information System (INIS)

Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E.

1991-09-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Ginna was selected as the eighth plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Ginna plant. 23 refs., 1 fig., 1 tab

Human Reliability Analysis: session summary

International Nuclear Information System (INIS)

Hall, R.E.

1985-01-01

The use of Human Reliability Analysis (HRA) to identify and resolve human factors issues has significantly increased over the past two years. Today, utilities, research institutions, consulting firms, and the regulatory agency have found a common application of HRA tools and Probabilistic Risk Assessment (PRA). The ''1985 IEEE Third Conference on Human Factors and Power Plants'' devoted three sessions to the discussion of these applications and a review of the insights so gained. This paper summarizes the three sessions and presents those common conclusions that were discussed during the meeting. The paper concludes that session participants supported the use of an adequately documented ''living PRA'' to address human factors issues in design and procedural changes, regulatory compliance, and training and that the techniques can produce cost effective qualitative results that are complementary to more classical human factors methods

Advances in human reliability analysis in Mexico

International Nuclear Information System (INIS)

Nelson, Pamela F.; Gonzalez C, M.; Ruiz S, T.; Guillen M, D.; Contreras V, A.

2010-10-01

Human Reliability Analysis (HRA) is a very important part of Probabilistic Risk Analysis (PRA), and constant work is dedicated to improving methods, guidance and data in order to approach realism in the results as well as looking for ways to use these to reduce accident frequency at plants. Further, in order to advance in these areas, several HRA studies are being performed globally. Mexico has participated in the International HRA Empirical study with the objective of -benchmarking- HRA methods by comparing HRA predictions to actual crew performance in a simulator, as well as in the empirical study on a US nuclear power plant currently in progress. The focus of the first study was the development of an understanding of how methods are applied by various analysts, and characterize the methods for their capability to guide the analysts to identify potential human failures, and associated causes and performance shaping factors. The HRA benchmarking study has been performed by using the Halden simulator, 14 European crews, and 15 HRA equipment s (NRC, EPRI, and foreign HRA equipment s using different HRA methods). This effort in Mexico is reflected through the work being performed on updating the Laguna Verde PRA to comply with the ASME PRA standard. In order to be considered an HRA with technical adequacy, that is, be considered as a capability category II, for risk-informed applications, the methodology used for the HRA in the original PRA is not considered sufficiently detailed, and the methodology had to upgraded. The HCR/CBDT/THERP method was chosen, since this is used in many nuclear plants with similar design. The HRA update includes identification and evaluation of human errors that can occur during testing and maintenance, as well as human errors that can occur during an accident using the Emergency Operating Procedures. The review of procedures for maintenance, surveillance and operation is a necessary step in HRA and provides insight into the possible

Perceptions of LWR risk for decision making

International Nuclear Information System (INIS)

Young, J.; Asselin, S.

1984-01-01

The Industry Degraded Core (IDCOR) Program was designed to develop a comprehensive, technically sound position on the issues related to potential accidents in light water reactors. One of the goals is to acquire knowledge and data so that a more realistic approach to the problem is possible. Some of the IDCOR tasks develop information in a Probabilistic Risk Assessment (PRA) framework. The PRA approach is structured upon reliability characteristics for individual components, such as pumps, valves and relays, which can be used to predict the frequency of system failures. System failure combinations can then be used to predict the probability of undesirable plant response to given initiating events. The IDCOR PRA tasks provide a significant amount of information related to the response of the plant to severe accidents. This information has been derived in a logical and consistent manner and so provides a coherent and rational basis for decision-making

Improved Monte Carlo Method for PSA Uncertainty Analysis

International Nuclear Information System (INIS)

Choi, Jongsoo

2016-01-01

The treatment of uncertainty is an important issue for regulatory decisions. Uncertainties exist from knowledge limitations. A probabilistic approach has exposed some of these limitations and provided a framework to assess their significance and assist in developing a strategy to accommodate them in the regulatory process. The uncertainty analysis (UA) is usually based on the Monte Carlo method. This paper proposes a Monte Carlo UA approach to calculate the mean risk metrics accounting for the SOKC between basic events (including CCFs) using efficient random number generators and to meet Capability Category III of the ASME/ANS PRA standard. Audit calculation is needed in PSA regulatory reviews of uncertainty analysis results submitted for licensing. The proposed Monte Carlo UA approach provides a high degree of confidence in PSA reviews. All PSA needs accounting for the SOKC between event probabilities to meet the ASME/ANS PRA standard

Improved Monte Carlo Method for PSA Uncertainty Analysis

Energy Technology Data Exchange (ETDEWEB)

Choi, Jongsoo [Korea Institute of Nuclear Safety, Daejeon (Korea, Republic of)

2016-10-15

The treatment of uncertainty is an important issue for regulatory decisions. Uncertainties exist from knowledge limitations. A probabilistic approach has exposed some of these limitations and provided a framework to assess their significance and assist in developing a strategy to accommodate them in the regulatory process. The uncertainty analysis (UA) is usually based on the Monte Carlo method. This paper proposes a Monte Carlo UA approach to calculate the mean risk metrics accounting for the SOKC between basic events (including CCFs) using efficient random number generators and to meet Capability Category III of the ASME/ANS PRA standard. Audit calculation is needed in PSA regulatory reviews of uncertainty analysis results submitted for licensing. The proposed Monte Carlo UA approach provides a high degree of confidence in PSA reviews. All PSA needs accounting for the SOKC between event probabilities to meet the ASME/ANS PRA standard.

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

Toward risk assessment 2.0: Safety supervisory control and model-based hazard monitoring for risk-informed safety interventions

International Nuclear Information System (INIS)

Favarò, Francesca M.; Saleh, Joseph H.

2016-01-01

Probabilistic Risk Assessment (PRA) is a staple in the engineering risk community, and it has become to some extent synonymous with the entire quantitative risk assessment undertaking. Limitations of PRA continue to occupy researchers, and workarounds are often proposed. After a brief review of this literature, we propose to address some of PRA's limitations by developing a novel framework and analytical tools for model-based system safety, or safety supervisory control, to guide safety interventions and support a dynamic approach to risk assessment and accident prevention. Our work shifts the emphasis from the pervading probabilistic mindset in risk assessment toward the notions of danger indices and hazard temporal contingency. The framework and tools here developed are grounded in Control Theory and make use of the state-space formalism in modeling dynamical systems. We show that the use of state variables enables the definition of metrics for accident escalation, termed hazard levels or danger indices, which measure the “proximity” of the system state to adverse events, and we illustrate the development of such indices. Monitoring of the hazard levels provides diagnostic information to support both on-line and off-line safety interventions. For example, we show how the application of the proposed tools to a rejected takeoff scenario provides new insight to support pilots’ go/no-go decisions. Furthermore, we augment the traditional state-space equations with a hazard equation and use the latter to estimate the times at which critical thresholds for the hazard level are (b)reached. This estimation process provides important prognostic information and produces a proxy for a time-to-accident metric or advance notice for an impending adverse event. The ability to estimate these two hazard coordinates, danger index and time-to-accident, offers many possibilities for informing system control strategies and improving accident prevention and risk mitigation

Risk-informed regulation

International Nuclear Information System (INIS)

Hoffman, D.R.

2003-01-01

In assessing safety for nuclear facilities, regulators have traditionally used a deterministic approach. New techniques for assessing nuclear or radiological risks make it possible for regulators to incorporate risk insights into their regulations. By 'risk-informing' the regulatory processes, independent bodies tasked with protecting the health and safety of the public can focus on those design and operational issues most important to safety. Such an approach is a move away from prescriptive regulations that were based on conservative engineering judgments toward regulations focused on issues that contribute significantly to safety. Despite the availability of probabilistic risk assessment (PRA) tools, organisations often struggle with how to best use this capability. Most international regulations are still based largely on deterministic analyses that were developed without the benefit of quantitative or measurable estimates of risk. PRA considers issues of risk in a more comprehensive manner by examining a wider spectrum of initiating events and their frequency, and considers the likelihood of events in a rigorous and comprehensive manner. In some countries, nuclear regulators are actively moving toward increasing the use of risk insights in a variety of strategic arenas, including risk-informed technical specifications (operating limits and conditions), in-service inspection and testing, programs, and assessment and enforcement actions. A risk-informed approach enhances the traditional deterministic approach by explicitly considering a broader range of safety challenges, focusing resources on the basis of risk significance, considering a broader range of counter measures to mitigate challenges, and explicitly identifying and quantifying uncertainties in analyses. (author)

Auxiliary feedwater system risk-based inspection guide for the Palo Verde Nuclear Power Plant

International Nuclear Information System (INIS)

Bumgardner, J.D.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.; Sloan, J.A.

1993-02-01

In a study sponsored by the US Nuclear Regulatory Commission (NRC), Pacific Northwest Laboratory has developed and applied a methodology for deriving plant-specific risk-based inspection guidance for the auxiliary feedwater (AFW) system at pressurized water reactors that have not undergone probabilistic risk assessment (PRA). This methodology uses existing PRA results and plant operating experience information. Existing PRA-based inspection guidance information recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. Palo Verde was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Palo Verde plants