National Library of Energy BETA

Sample records for usive pra ctices

  1. PRA and Risk Informed Analysis

    SciTech Connect (OSTI)

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

    2006-01-01

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

  2. Linkage of PRA models. Phase 1, Results

    SciTech Connect (OSTI)

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

    1995-12-01

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

  3. Probabilistic risk assessment course documentation. Volume 1: PRA fundamentals

    SciTech Connect (OSTI)

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

    1985-08-01

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

  4. Attendees- December 2014 P&RA Technical Exchange Meeting

    Office of Energy Efficiency and Renewable Energy (EERE)

    Attendees to the Performance & Risk Assessment Community of Practice (P&RA) Technical Exchange Meeting held in Las Vegas, Nevada on December 11 & 12, 2014.

  5. Performance & Risk Assessment Community of Practice (P&RA CoP...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Performance & Risk Assessment Community of Practice (P&RA CoP) Performance & Risk Assessment Community of Practice (P&RA CoP) P&RA CoP's Technical Exchange Meeting held on December ...

  6. Performance & Risk Assessment Community of Practice (P&RA CoP...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Performance & Risk Assessment Community of Practice (P&RA CoP) Technical Exchanges Performance & Risk Assessment Community of Practice (P&RA CoP) Technical Exchanges PA CoP has ...

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

    SciTech Connect (OSTI)

    none,

    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.

  8. Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) Demonstration

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01

    A key area of the Advanced Small Modular Reactor (SMR) Probabilistic Risk Assessment (PRA) strategy is the development of methodologies and tools that will be used to predict the safety, security, safeguards, performance, and deployment viability of SMRs. The goal of the SMR PRA activity will be to develop quantitative methods and tools and the associated analysis framework for assessing a variety of risks. Development and implementation of SMR-focused safety assessment methods may require new analytic methods or adaptation of traditional methods to the advanced design and operational features of SMRs. We will need to move beyond the current limitations such as static, logic-based models in order to provide more integrated, scenario-based models based upon predictive modeling which are tied to causal factors. The development of SMR-specific safety models for margin determination will provide a safety case that describes potential accidents, design options (including postulated controls), and supports licensing activities by providing a technical basis for the safety envelope. This report documents the progress that was made to implement the PRA framework, specifically by way of demonstration of an advanced 3D approach to representing, quantifying and understanding flooding risks to a nuclear power plant.

  9. List of Topics for Interagency Performance & Risk Assessment Community of Practice (P&RA CoP) Discussion

    Office of Energy Efficiency and Renewable Energy (EERE)

    List of Topics for Interagency Performance & Risk Assessment Community of Practice (P&RA CoP) Discussion

  10. DYNAMIC AND CLASSICAL PRA: A BWR SBO CASE COMPARISON

    SciTech Connect (OSTI)

    Mandelli, Diego; Smith, Curtis L; Ma, Zhegang

    2011-07-01

    As part of the Light-Water Sustainability Program (LWRS), the purpose of the Risk Informed Safety Margin Characterization (RISMC) Pathway research and development is to support plant decisions for risk-informed margin management with the aim to improve economics, reliability, and sustain the safety of current NPPs. In this paper, we describe the RISMC analysis process illustrating how mechanistic (i.e., dynamic system simulators) and probabilistic (stochastic sampling strategies) approaches are combined in a dynamic PRA fashion in order to estimate safety margins. We use the scenario of a “station blackout” (SBO) wherein offsite power and onsite power are lost, thereby causing a challenge to plant safety systems. We describe the RISMC approach, illustrate the station blackout modeling, and compare this with traditional risk analysis modeling for this type of accident scenario. In the RISMC approach the dataset obtained consists of set of simulation runs (performed by using codes such as RELAP5/3D) where timing and ordering of events is changed accordingly to the stochastic sampling strategy adopted. On the other side, classical PRA methods, which are based on event-tree (FT) and fault-tree (FT) structures, generate minimal cut sets and probability values associated to each ET branch. The comparison of the classical and RISMC approaches is performed not only in terms of overall core damage probability but also considering statistical differences in the actual sequence of events. The outcome of this comparison analysis shows similarities and dissimilarities between the approaches but also highlights the greater amount of information that can be generated by using the RISMC approach.

  11. Biosketches of Speakers- P&RA CoP December 2014 Technical Exchange Meeting

    Office of Energy Efficiency and Renewable Energy (EERE)

    Bio-sketches of Speakers from the Performance & Risk Assessment Community of Practice (P&RA) Technical Exchange Meeting held in Las Vegas, Nevada on December 11 & 12, 2014.

  12. Level 1 Tornado PRA for the High Flux Beam Reactor

    SciTech Connect (OSTI)

    Bozoki, G.E.; Conrad, C.S.

    1994-05-01

    This report describes a risk analysis primarily directed at providing an estimate for the frequency of tornado induced damage to the core of the High Flux Beam Reactor (HFBR), and thus it constitutes a Level 1 Probabilistic Risk Assessment (PRA) covering tornado induced accident sequences. The basic methodology of the risk analysis was to develop a ``tornado specific`` plant logic model that integrates the internal random hardware failures with failures caused externally by the tornado strike and includes operator errors worsened by the tornado modified environment. The tornado hazard frequency, as well as earlier prepared structural and equipment fragility data, were used as input data to the model. To keep modeling/calculational complexity as simple as reasonable a ``bounding`` type, slightly conservative, approach was applied. By a thorough screening process a single dominant initiating event was selected as a representative initiator, defined as: ``Tornado Induced Loss of Offsite Power.`` The frequency of this initiator was determined to be 6.37E-5/year. The safety response of the HFBR facility resulted in a total Conditional Core Damage Probability of .621. Thus, the point estimate of the HFBR`s Tornado Induced Core Damage Frequency (CDF) was found to be: (CDF){sub Tornado} = 3.96E-5/year. This value represents only 7.8% of the internal CDF and thus is considered to be a small contribution to the overall facility risk expressed in terms of total Core Damage Frequency. In addition to providing the estimate of (CDF){sub Tornado}, the report documents, the relative importance of various tornado induced system, component, and operator failures that contribute most to (CDF){sub Tornado}.

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

    SciTech Connect (OSTI)

    Curtis Smith

    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.

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Luckas, W.J.; Barriere, M.T.; Brown, W.S. ); Wreathall, J. and Co., Dublin, OH ); Cooper, S.E. )

    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).

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

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    Not Available

    2004-09-01

    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.

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

    SciTech Connect (OSTI)

    Tom Elicson; Bentley Harwood; Jim Bouchard; Heather Lucek

    2011-03-01

    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.

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

    SciTech Connect (OSTI)

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

    2011-03-01

    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.

  20. Comparing Simulation Results with Traditional PRA Model on a Boiling Water Reactor Station Blackout Case Study

    SciTech Connect (OSTI)

    Zhegang Ma; Diego Mandelli; Curtis Smith

    2011-07-01

    A previous study used RELAP and RAVEN to conduct a boiling water reactor station black-out (SBO) case study in a simulation based environment to show the capabilities of the risk-informed safety margin characterization methodology. This report compares the RELAP/RAVEN simulation results with traditional PRA model results. The RELAP/RAVEN simulation run results were reviewed for their input parameters and output results. The input parameters for each simulation run include various timing information such as diesel generator or offsite power recovery time, Safety Relief Valve stuck open time, High Pressure Core Injection or Reactor Core Isolation Cooling fail to run time, extended core cooling operation time, depressurization delay time, and firewater injection time. The output results include the maximum fuel clad temperature, the outcome, and the simulation end time. A traditional SBO PRA model in this report contains four event trees that are linked together with the transferring feature in SAPHIRE software. Unlike the usual Level 1 PRA quantification process in which only core damage sequences are quantified, this report quantifies all SBO sequences, whether they are core damage sequences or success (i.e., non core damage) sequences, in order to provide a full comparison with the simulation results. Three different approaches were used to solve event tree top events and quantify the SBO sequences: W process flag, default process flag without proper adjustment, and default process flag with adjustment to account for the success branch probabilities. Without post-processing, the first two approaches yield incorrect results with a total conditional probability greater than 1.0. The last approach accounts for the success branch probabilities and provides correct conditional sequence probabilities that are to be used for comparison. To better compare the results from the PRA model and the simulation runs, a simplified SBO event tree was developed with only four top

  1. Review results of a BWR standard plant PRA and an assessment of potential benefits from design modifications

    SciTech Connect (OSTI)

    Shiu, K.; Hanan, N.; Rubin, M.

    1985-01-01

    Brookhaven National Laboratory (BNL) has participated in the review of the GESSAR II Standard Boiling Water Reactor (BWR) Plant probabilistic risk assessment (PRA). One major objective of this review was to utilize the PRA as a tool for investigation of the relative benefits available for incorporation of various proposed modifications to the baseline design. This paper presents the findings of the BNL review and assessment of the impact upon core damage frequency from two suggested design modifications. This work was restricted to consideration of interal events only. Review results indicated that the point estimate core damage frequency of the GESSAR II plant is equal to 2.2 x 10/sup -5//reactor-year for a plant site located within the Mid-Atlantic Area Council Grid (MAAC) and 3.8 x 10/sup -5//reactor-year if the national average loss of offsite power initiator frequency is used.

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

    SciTech Connect (OSTI)

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

    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).

  3. FOIA-PRA Doc.pdf

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

  4. Loss of spent fuel pool cooling PRA: Model and results

    SciTech Connect (OSTI)

    Siu, N.; Khericha, S.; Conroy, S.; Beck, S.; Blackman, H.

    1996-09-01

    This letter report documents models for quantifying the likelihood of loss of spent fuel pool cooling; models for identifying post-boiling scenarios that lead to core damage; qualitative and quantitative results generated for a selected plant that account for plant design and operational practices; a comparison of these results and those generated from earlier studies; and a review of available data on spent fuel pool accidents. The results of this study show that for a representative two-unit boiling water reactor, the annual probability of spent fuel pool boiling is 5 {times} 10{sup {minus}5} and the annual probability of flooding associated with loss of spent fuel pool cooling scenarios is 1 {times} 10{sup {minus}3}. Qualitative arguments are provided to show that the likelihood of core damage due to spent fuel pool boiling accidents is low for most US commercial nuclear power plants. It is also shown that, depending on the design characteristics of a given plant, the likelihood of either: (a) core damage due to spent fuel pool-associated flooding, or (b) spent fuel damage due to pool dryout, may not be negligible.

  5. Enhanced Fire Events Database to Support Fire PRA

    SciTech Connect (OSTI)

    Patrick Baranowsky; Ken Canavan; Shawn St. Germain

    2010-06-01

    Abstract: This paper provides a description of the updated and enhanced Fire Events Data Base (FEDB) developed by the Electric Power Research Institute (EPRI) in cooperation with the U.S. Nuclear Regulatory Commission (NRC). The FEDB is the principal source of fire incident operational data for use in fire PRAs. It provides a comprehensive and consolidated source of fire incident information for nuclear power plants operating in the U.S. The database classification scheme identifies important attributes of fire incidents to characterize their nature, causal factors, and severity consistent with available data. The database provides sufficient detail to delineate important plant specific attributes of the incidents to the extent practical. A significant enhancement to the updated FEDB is the reorganization and refinement of the database structure and data fields and fire characterization details added to more rigorously capture the nature and magnitude of the fire and damage to the ignition source and nearby equipment and structures

  6. SPACE PROPULSION SYSTEM PHASED-MISSION PROBABILITY ANALYSIS USING CONVENTIONAL PRA METHODS

    SciTech Connect (OSTI)

    Curtis Smith; James Knudsen

    2006-05-01

    As part of a series of papers on the topic of advance probabilistic methods, a benchmark phased-mission problem has been suggested. This problem consists of modeling a space mission using an ion propulsion system, where the mission consists of seven mission phases. The mission requires that the propulsion operate for several phases, where the configuration changes as a function of phase. The ion propulsion system itself consists of five thruster assemblies and a single propellant supply, where each thruster assembly has one propulsion power unit and two ion engines. In this paper, we evaluate the probability of mission failure using the conventional methodology of event tree/fault tree analysis. The event tree and fault trees are developed and analyzed using Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE). While the benchmark problem is nominally a "dynamic" problem, in our analysis the mission phases are modeled in a single event tree to show the progression from one phase to the next. The propulsion system is modeled in fault trees to account for the operation; or in this case, the failure of the system. Specifically, the propulsion system is decomposed into each of the five thruster assemblies and fed into the appropriate N-out-of-M gate to evaluate mission failure. A separate fault tree for the propulsion system is developed to account for the different success criteria of each mission phase. Common-cause failure modeling is treated using traditional (i.e., parametrically) methods. As part of this paper, we discuss the overall results in addition to the positive and negative aspects of modeling dynamic situations with non-dynamic modeling techniques. One insight from the use of this conventional method for analyzing the benchmark problem is that it requires significant manual manipulation to the fault trees and how they are linked into the event tree. The conventional method also requires editing the resultant cut sets to obtain the correct results. While conventional methods may be used to evaluate a dynamic system like that in the benchmark, the level of effort required may preclude its use on real-world problems.

  7. A methodology for generating dynamic accident progression event trees for level-2 PRA

    SciTech Connect (OSTI)

    Hakobyan, A.; Denning, R.; Aldemir, T. [Ohio State Univ., Nuclear Engineering Program, 650 Ackerman Road, Columbus, OH 43202 (United States); Dunagan, S.; Kunsman, D. [Sandia National Laboratory, Albuquerque, NM 87185 (United States)

    2006-07-01

    Currently, the development and analysis of Accident Progression Event Trees (APETs) are performed in a manner that is computationally time consuming, difficult to reproduce and also can be phenomenologically inconsistent. A software tool (ADAPT) is described for automated APET generation using the concept of dynamic event trees. The tool determines the branching times from a severe accident analysis code based on user specified criteria for branching. It assigns user specified probabilities to every branch, tracks the total branch probability, and truncates branches based on the given pruning/truncation rules to avoid an unmanageable number of scenarios. While the software tool could be applied to any systems analysis code, the MELCOR code is used for this illustration. A case study is presented involving station blackout with the loss of auxiliary feedwater system for a pressurized water reactor. (authors)

  8. U.S. NRC CONFIRMATORY LEVEL 1 PRA SUCCESS CRITERIA ACTIVITIES

    SciTech Connect (OSTI)

    Donald Helton; Hossein Esmaili; Robert Buell

    2011-03-01

    The U.S. Nuclear Regulatory Commission’s standardized plant analysis risk (SPAR) models are used to support a number of risk-informed initiatives. The fidelity and realism of these models are ensured through a number of processes including cross-comparison with industry models, review and use by a wide range of technical experts, and confirmatory analysis. This paper will describe a key activity in the latter arena. Specifically, this paper will describe MELCOR analyses performed to augment the technical basis for confirming or modifying specific success criteria of interest. The analyses that will be summarized provide the basis for confirming or changing success criteria in a specific 3-loop pressurized-water reactor and a Mark-I boiling-water reactor. Initiators that have been analyzed include loss-of-coolant accidents, loss of main feedwater, spontaneous steam generator tube rupture, inadvertent opening of a relief valve at power, and station blackout. For each initiator, specific aspects of the accident evolution are investigated via a targeted set of calculations (3 to 22 distinct accident analyses per initiator). Further evaluation is ongoing to extend the analyses’ conclusions to similar plants (where appropriate), with consideration of design and modeling differences on a scenario-by-scenario basis. This paper will also describe future plans.

  9. Microsoft Word - PRA CoP Techncial Exchange Draft Agenda 2015...

    Office of Environmental Management (EM)

    Building 10:30 am - 11:15 pm Approaches for Uncertainty Quantification and Sensitivity Analysis, Dr. Matt Kozak (INTERA) 11:15 am - 12:00 pm Use of Performance and Risk...

  10. Microsoft PowerPoint - P&RA CoP EPA optimization Biggs final...

    Office of Environmental Management (EM)

    ...Biggs703-823-3081biggs.kirby@epa.gov * Regional management involved in optimization * ... * Historical information and data * Geology, hydrogeology, chemistry, operations * ...

  11. Validation and verification plan for safety and PRA codes. Revision 1

    SciTech Connect (OSTI)

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

    1991-04-01

    This report discusses a verification and validation (V&V) plan for computer codes used for safety analysis and probabilistic risk assessment calculations. The present plan fulfills the commitments by Westinghouse Savannah River Company (WSRC) to the Department of Energy Savannah River Office (DOE-SRO) to bring the essential safety analysis and probabilistic risk assessment codes in compliance with verification and validation requirements.

  12. P&RA CoP TE Mtg Attendees.xlsx

    Office of Environmental Management (EM)

    EM 74 Humphries, Ginger SRNS 75 Krenzien, Susan Navarro 76 Mallick, Pramod DOE EM 77 Moore, Beth DOE EM 78 Parks, Leah 79 Proehl, Gerhard IAEA 80 Richards, Jon EPA Region 4 81...

  13. Microsoft Word - P&RA CoP Techncial Exchange Final Agenda

    Office of Environmental Management (EM)

    Agenda Interagency Steering Committee on Performance and Risk Assessment Community of Practice Annual Technical Exchange Meeting December 15 and 16, 2015 Washington State ...

  14. P&RA CoP TE Mtg Attendees List _2014-12-23.xlsx

    Office of Environmental Management (EM)

    AgencyCompany Affliation 1 George Alexander NRC 2 Alaa Aly INTERA 3 Bob Andrews INTERA 4 Cynthia Barr NRC 5 Debbie Barr DOE LM 6 Craig Benson University of Wisconsin-Madison 7 ...

  15. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    York, GMT-04:00) | 1 hr 30 min Join WebEx meeting Meeting number: 998 683 367 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: ...

  16. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    Time (New York, GMT-05:00) Meeting Number: 993 330 453 Meeting Password: (This meeting does not require a password.) ---... To ...

  17. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    ...j.php?MTIDm67be2f2b90f2b04150c50213e0dcd237 Meeting number: 992 479 747 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: ...

  18. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    ...j.php?MTIDm64294cde2069b1706de2237fd87b0607 Meeting number: 997 009 337 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: ...

  19. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    Meeting number: 995 136 661 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: 995 136 661 Add this meeting to your calendar. ...

  20. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    ...j.php?MTIDm9b93ed24d0ae38ab9e7b0fb59fd5dab7 Meeting number: 998 033 386 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: ...

  1. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    Meeting number: 998 420 290 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: 998 420 290 Add this meeting to your calendar. ...

  2. Microsoft Word - 2015-05-20 PRA CoP Webinar Agenda

    Office of Environmental Management (EM)

    Meeting number: 998 420 290 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: 998 420 290 Add this meeting to your calendar. ...

  3. Microsoft Word - 2013-12-12 P&RA CoP Webinar_121113

    Office of Environmental Management (EM)

    ...j.php?MTIDm353fba8bab906abcb4b2af89d0ef05c0 Meeting number: 990 595 675 Meeting password: Meeting1 Join by phone 1-650-479-3208 Call-in toll number (USCanada) Access code: ...

  4. Microsoft Word - 2014-06-03 P&RA CoP Webinar

    Office of Environmental Management (EM)

    Time (New York, GMT-04:00) Meeting Number: 991 895 794 Meeting Password: (This meeting does not require a password.) ---... To ...

  5. Performance & Risk Assessment Community of Practice (P&RA CoP)

    Broader source: Energy.gov [DOE]

    Performance assessments (PAs) and risk assessments (RAs) evaluate the impact of a proposed remedial action on human health and the environment, and provide a demonstration of compliance and important technical inputs to meet regulatory requirements for: 1) waste form development and implementation; 2) tank closure activities; 3) waste site closure activities (e.g., cribs and trenches); 4) in-situ decontamination and decommissioning; 5) soil and groundwater remediation; and 6) management of disposal facilities (e.g., land-fills or near surface disposal facilities). The PAs and RAs or P&RAs become public documents upon completion. As such, the Department of Energy (DOE) needs to ensure that P&RAs continue to be performed and documented consistently and to high standards.

  6. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated,more » where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.« less

  7. Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

    Broader source: Energy.gov [DOE]

    During the 580th meeting of the Advisory Committee on Reactor Safeguards (ACRS), February10-12, 2011, we reviewed the staff’s white paper, “A Comparison of Integrated Safety Analysisand...

  8. Analyzing simulation-based PRA data through traditional and topological clustering: A BWR station blackout case study

    SciTech Connect (OSTI)

    Maljovec, D.; Liu, S.; Wang, B.; Mandelli, D.; Bremer, P. -T.; Pascucci, V.; Smith, C.

    2015-07-14

    Here, dynamic probabilistic risk assessment (DPRA) methodologies couple system simulator codes (e.g., RELAP and MELCOR) with simulation controller codes (e.g., RAVEN and ADAPT). Whereas system simulator codes model system dynamics deterministically, simulation controller codes introduce both deterministic (e.g., system control logic and operating procedures) and stochastic (e.g., component failures and parameter uncertainties) elements into the simulation. Typically, a DPRA is performed by sampling values of a set of parameters and simulating the system behavior for that specific set of parameter values. For complex systems, a major challenge in using DPRA methodologies is to analyze the large number of scenarios generated, where clustering techniques are typically employed to better organize and interpret the data. In this paper, we focus on the analysis of two nuclear simulation datasets that are part of the risk-informed safety margin characterization (RISMC) boiling water reactor (BWR) station blackout (SBO) case study. We provide the domain experts a software tool that encodes traditional and topological clustering techniques within an interactive analysis and visualization environment, for understanding the structures of such high-dimensional nuclear simulation datasets. We demonstrate through our case study that both types of clustering techniques complement each other for enhanced structural understanding of the data.

  9. Interagency Performance and Risk Assessment Community of Practice...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Interagency Performance and Risk Assessment Community of Practice (P&RA CoP) Charter Interagency Performance and Risk Assessment Community of Practice (P&RA CoP) Charter Charter ...

  10. List of Topics for Interagency Performance & Risk Assessment...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    & Risk Assessment Community of Practice (P&RA CoP) Discussion List of Topics for Interagency Performance & Risk Assessment Community of Practice (P&RA CoP) Discussion List of ...

  11. 7th Workshop on Risk Informed Regulation and Safety Culture ...

    Energy Savers [EERE]

    their PRA. However, Qinshan will hire two U.S. PRA firms to develop its "Generation Risk Analysis" model. This summer, a team of Qinshan PSA personnel will visit South Texas...

  12. The 10,000-year debate

    SciTech Connect (OSTI)

    Wilson, J.R.

    1996-08-01

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

  13. Status Updates on the Performance and Risk Assessment Community...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Status Updates on the Performance and Risk Assessment Community of Practice (P&RA CoP) Ming Zhu, Ph.D., PE, PMP Chair of P&RA CoP P&RA CoP Technical Exchange Meeting Las Vegas, NV ...

  14. Radiation transport and energetics of laser-driven half-hohlraums at the National Ignition Facility

    SciTech Connect (OSTI)

    Moore, A. S.; Cooper, A. B.R.; Schneider, M. B.; MacLaren, S.; Graham, P.; Lu, K.; Seugling, R.; Satcher, J.; Klingmann, J.; Comley, A. J.; Marrs, R.; May, M.; Widmann, K.; Glendinning, G.; Castor, J.; Sain, J.; Back, C. A.; Hund, J.; Baker, K.; Hsing, W. W.; Foster, J.; Young, B.; Young, P.

    2014-06-01

    Experiments that characterize and develop a high energy-density half-hohlraum platform for use in bench-marking radiation hydrodynamics models have been conducted at the National Ignition Facility (NIF). Results from the experiments are used to quantitatively compare with simulations of the radiation transported through an evolving plasma density structure, colloquially known as an N-wave. A half-hohlraum is heated by 80 NIF beams to a temperature of 240 eV. This creates a subsonic di usive Marshak wave which propagates into a high atomic number Ta2O5 aerogel. The subsequent radiation transport through the aerogel and through slots cut into the aerogel layer is investigated. We describe a set of experiments that test the hohlraum performance and report on a range

  15. A review of NRC staff uses of probabilistic risk assessment

    SciTech Connect (OSTI)

    Not Available

    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.

  16. Information Collection Management | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Program - Sets forth Department of Energy (DOE) PRA inplementation requirements. ... Instructions - ICR OMB 83-I-Paperwork Reduction Act Submission Form OMB 83-D-Paperwork ...

  17. DOE-STD-1104 Acronyms

    Office of Environmental Management (EM)

    Position O Order PDSA Preliminary Documented Safety Analysis PRA Probabilistic Risk Assessment PSDR Preliminary Safety Design Report SBAA Safety Basis Approval Authority SBRT...

  18. NETL F 451.1/1-1, Categorical Exclusion Designation Form

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    FE0025387 PRA Multiple sites in AK Environmental Resources Management Alaska Inc. (ERM); Loundsbury & Associates, Inc.; Peak Oilfield Services Company, LLC; Maritime Helicopters...

  19. Microsoft Word - List of topics_2015-11-12

    Office of Environmental Management (EM)

    the 2014 P&RA CoP Technical Exchange Meeting: * Confidence building in performance and risk assessments models (Kirby Biggs, November 12, 2015) * Speciation and transport of...

  20. Statement of Intent NO. 2 between the US Department of Energy...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Site & Facility Restoration Deactivation & Decommissioning (D&D) P&RA Community of Practice Facility Engineering Soil & Groundwater Sustainability Program Management Safety ...

  1. Statement of Intent between US Department of Energy and the State...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Site & Facility Restoration Deactivation & Decommissioning (D&D) P&RA Community of Practice Facility Engineering Soil & Groundwater Sustainability Program Management Safety ...

  2. February 20, 2014 Webinar- Performance of Engineered Barriers: Lessons Learned

    Broader source: Energy.gov [DOE]

    P&RA CoP Webinar - 2/20/2014 - Performance of Engineered Barriers: Lessons Learned Craig H. Benson (University of Wisconsin-Madison/CRESP)

  3. Discussions

    Broader source: Energy.gov [DOE]

    Discussions - Presentations and closing discussions of the P&RA CoP December 11-12, 2014 Technical Exchange Meeting

  4. 2015 Annual Technical Exchange Meeting Presenters Biographical...

    Office of Environmental Management (EM)

    Sketches 2015 Annual Technical Exchange Meeting Presenters Biographical Sketches Interagency Performance & Risk Assessment Community of Practice (P&RA CoP) 2015 Annual ...

  5. Training for Records and Information Management

    Broader source: Energy.gov [DOE]

    Records Management Training:  NARA Records Management Training   NARA Targeted Assistance NARA Brochures Training Presentation:  Information Collection Requests/PRA (pdf)  

  6. Procedures for Obtaining OMB Clearance to Conduct a Survey

    SciTech Connect (OSTI)

    2009-01-18

    This appendix uses two flow charts (General Clearance Process and PRA Review Process) to provide a visual image of the OMB clearance process.

  7. Analysis of the Space Propulsion System Problem Using RAVEN ...

    Office of Scientific and Technical Information (OSTI)

    RAVEN (Reactor Analysis and Virtual control ENviroment) is a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities. ...

  8. Search for: All records | SciTech Connect

    Office of Scientific and Technical Information (OSTI)

    for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for ... We present a software tool that is designed to address these goals. We model a ...

  9. Exploratory Nuclear Reactor Safety Analysis and Visualization...

    Office of Scientific and Technical Information (OSTI)

    The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic ... than one failure mode, control loops, andor hardwareprocesssoftwarehuman interaction. ...

  10. November 12, 2015 Webinar- Implementing Optimization in the Superfund Program

    Broader source: Energy.gov [DOE]

    P&RA CoP Webinar - November 12, 2015 Webinar - Implementing Optimization in the Superfund Program, Mr. Kirby Biggs and Mr. Daniel Powell (EPA OSRTI).

  11. Permitted Mercury Storage Facility Notifications | Department...

    Office of Environmental Management (EM)

    submitted notificationcertification letters to DOE stating that they meet the ... Site & Facility Restoration Deactivation & Decommissioning (D&D) P&RA Community of ...

  12. Risk assessment of CST-7 proposed waste treatment and storage facilities Volume I: Limited-scope probabilistic risk assessment (PRA) of proposed CST-7 waste treatment & storage facilities. Volume II: Preliminary hazards analysis of proposed CST-7 waste storage & treatment facilities

    SciTech Connect (OSTI)

    Sasser, K.

    1994-06-01

    In FY 1993, the Los Alamos National Laboratory Waste Management Group [CST-7 (formerly EM-7)] requested the Probabilistic Risk and Hazards Analysis Group [TSA-11 (formerly N-6)] to conduct a study of the hazards associated with several CST-7 facilities. Among these facilities are the Hazardous Waste Treatment Facility (HWTF), the HWTF Drum Storage Building (DSB), and the Mixed Waste Receiving and Storage Facility (MWRSF), which are proposed for construction beginning in 1996. These facilities are needed to upgrade the Laboratory`s storage capability for hazardous and mixed wastes and to provide treatment capabilities for wastes in cases where offsite treatment is not available or desirable. These facilities will assist Los Alamos in complying with federal and state requlations.

  13. Augmenting Probabilistic Risk Assesment with Malevolent Initiators

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder

    2011-11-01

    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.

  14. System Analysis and Risk Assessment System.

    Energy Science and Technology Software Center (OSTI)

    2000-11-20

    Version 00 SARA4.16 is a program that allows the user to review the results of a Probabilistic Risk Assessment (PRA) and to perform limited sensitivity analysis on these results. This tool is intended to be used by a less technical oriented user and does not require the level of understanding of PRA concepts required by a full PRA analysis tool. With this program a user can review the information generated by a PRA analyst andmore » compare the results to those generated by making limited modifications to the data in the PRA. Also included in this program is the ability to graphically display the information stored in the database. This information includes event trees, fault trees, P&IDs and uncertainty distributions. SARA 4.16 is incorporated in the SAPHIRE 5.0 code package.« less

  15. Operational phase of inspection prioritization

    SciTech Connect (OSTI)

    Campbell, D.J.; Guthrie, V.H.; Flanagan, G.F.

    1985-01-01

    Inspectors must make many decisions on the allocation of their efforts. To date, these decisions have been made based upon their own judgment and guidance from inspection procedures. The goal of this paper is to provide PRA information as an additional aid to inspectors. A structured approach for relating PRA information to specific inspection decisions has been developed. The use of PRA information as an aid in optimal decision making (1) in response to the current plant status and (2) in the scheduling of effort over an extended period of time is considered. 21 figs.

  16. February 5, 2014 Webinar- The Cementitious Barriers Partnership Toolbox, Version 2.0

    Office of Energy Efficiency and Renewable Energy (EERE)

    P&RA CoP Webinar - February 5, 2014 - Tools and Capabilities of the Cementitious Barriers Partnership Toolbox, Version 2.0 David Kosson et al. (Vanderbilt University/CRESP)

  17. Quality Assurance for Performance Assessment Modeling

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  18. Models And Results Database System.

    Energy Science and Technology Software Center (OSTI)

    2001-03-27

    Version 00 MAR-D 4.16 is a program that is used primarily for Probabilistic Risk Assessment (PRA) data loading. This program defines a common relational database structure that is used by other PRA programs. This structure allows all of the software to access and manipulate data created by other software in the system without performing a lengthy conversion. The MAR-D program also provides the facilities for loading and unloading of PRA data from the relational databasemore » structure used to store the data to an ASCII format for interchange with other PRA software. The primary function of MAR-D is to create a data repository for NUREG-1150 and other permanent data by providing input, conversion, and output capabilities for data used by IRRAS, SARA, SETS and FRANTIC.« less

  19. Risk Analysis and Decision-Making Under Uncertainty: A Strategy...

    Office of Environmental Management (EM)

    Uncertainty Analysis and Parameter Estimation Since 2002 To view all the P&RA CoP ... Update of Hydrogen from Biomass - Determination of the Delivered Cost of Hydrogen: ...

  20. Risk Informing Environmental Cleanup Priorities

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  1. Liners and Covers: Field Performance & Life Expectancy | Department...

    Office of Environmental Management (EM)

    Meeting 11-12 December 2014 Las Vegas, Nevada, USA To view all the P&RA CoP 2014 ... an Alternative Landfill Cover at the Monticello, Utah, Uranium Mill Tailings Disposal Site

  2. Cementitious Barrier Partnership Program Update

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  3. February 23, 2016 Webinar- Multi-Criteria Decisional Analyses: Methodology and Case Studies

    Broader source: Energy.gov [DOE]

    Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - February 23, 2016 - Multi-Criteria Decisional Analyses: Methodology and Case Studies (Dr. Igor Linkov and Mr. Matthew Bates, U.S. Army Corps of Engineers).

  4. Scaling of Saltstone Disposal Facility Testing

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  5. MODARIA: Modelling and Data for Radiological Impact Assessment Context and Overview

    Office of Energy Efficiency and Renewable Energy (EERE)

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  6. Hanford Site Waste Management Area C Performance Assessment

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  7. Status of SRS Liquid Waste Performance Assessment Program

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  8. May 20, 2015 Webinar - Guidance for Conducting Technical Analyses...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - May 20, 2015 - Guidance for Conducting Technical Analyses for 10 CFR Part 61 by Mr. Chris Grossman (NRC) ...

  9. E-Area Performance Assessment

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  10. Summary of NWTRB Deep Borehole Disposal Workshop

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  11. Deep Borehole Disposal (DBD) Performance Assessment

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  12. December 12, 2013 Webinar- The Use of Graded Approach in Hanford Vadose Zone Modeling

    Broader source: Energy.gov [DOE]

    P&RA CoP Webinar - Dec. 12, 2013 - Alaa Aly (INTERA) & Dib Goswami (Washington State Ecology), “The Use of Graded Approach in Hanford Vadose Zone Modeling”

  13. Microsoft PowerPoint - CoP Webinar Jan 2016

    Office of Environmental Management (EM)

    Borehole Disposal of Spent Sources (BOSS) Matthew W. Kozak INTERA, Inc. Denver, CO Interagency P&RA Community of Practice Webinar January 28, 2016 BOSS Program IAEA program to ...

  14. Using Performance Assessments to Focus Research & Development Activities

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  15. Analysis of Environmental Monitoring Data Following Site Closure...

    Office of Environmental Management (EM)

    and Risk Assessment Community of Practice Annual Technical Exchange Meeting Richard Bush, Program Manager December 11, 2014 To view all the P&RA CoP 2014 Technical Exchange...

  16. February 23, 2016 Webinar - Multi-Criteria Decisional Analyses...

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Analysis: Environmental Applications and Case Studies", by Igor Linkov & Emily Moberg, CRC Press, 2011 Agenda & Webinar Instructions - February 23, 2016 - P&RA CoP Webinar (169.55 ...

  17. Performance Assessment of the Portsmouth On-Site Waste Disposal Facility

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  18. Information Management and Supporting Documentation

    Broader source: Energy.gov [DOE]

    The Paperwork Reduction Act (PRA) of 1995 requires each Federal agency to seek and obtain approval from the Office of Management and Budget (OMB) before undertaking a collection of information...

  19. The Hanford Site-Wide Risk Review Project | Department of Energy

    Office of Environmental Management (EM)

    December 11, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation PDF icon The Hanford Site-Wide Risk Review Project More...

  20. Highlights from a Workshop Series: Best Practices for Risk-Informed...

    Office of Environmental Management (EM)

    Technical Exchange Meeting To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation PDF icon Highlights from a Workshop Series: Best...

  1. Hanford Site Waste Management Area C Performance Assessment ...

    Office of Environmental Management (EM)

    Exchange December 11-12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation - Part 1 Video Presentation - Part 2 PDF icon Hanford...

  2. On Performance of Covers and Liners In Performance Assessments...

    Office of Environmental Management (EM)

    December 11 and 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation PDF icon On Performance of Covers and Liners In...

  3. fe0025387-Petrotechnical-Resources | netl.doe.gov

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    ... DOE Contribution: 1,146,132 Performer Contribution: NA Contact Information: NETL - Skip Pratt (skip.pratt@netl.doe.gov or 304-285-4396) PRA - Robert Hunter (rhunter@petroak.com or ...

  4. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91). Final report

    SciTech Connect (OSTI)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. [eds.

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  5. Code development incorporating environmental, safety, and economic aspects of fusion reactors (FY 89--91)

    SciTech Connect (OSTI)

    Ho, S.K.; Fowler, T.K.; Holdren, J.P. (eds.)

    1991-11-01

    This report discusses the following aspects of Fusion reactors.: Activation Analysis; Tritium Inventory; Environmental and Safety Indices and Their Graphical Representation; Probabilistic Risk Assessment (PRA) and Decision Analysis; Plasma Burn Control -- Application to ITER; and Other Applications.

  6. Approaches for Uncertainty Quantification and Sensitivity Analysis

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  7. Managing Uncertainty and Demonstrating Compliance

    Broader source: Energy.gov [DOE]

    Presentation from the 2015 Annual Performance and Risk Assessment (P&RA) Community of Practice (CoP) Technical Exchange Meeting held in Richland, Washington on December 15-16, 2015.

  8. Paperwork Reduction Act for Surveys and User Research | Department...

    Office of Environmental Management (EM)

    If you don't know whether your survey requires OMB approval, contact the Web Usability ... Contact the Web Usability Coordinator and ask for the Paperwork Reduction Act (PRA) Fast ...

  9. PAPERWORK REDUCTION ACT OF 1995

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    PAPERWORK REDUCTION ACT OF 1995 U. S. DEPARTMENT OF ENERGY INFORMATION COLLECTION MANAGEMENT PROGRAM Chris Rouleau, PRA Officer Records Management Division Office of the Associate Chief Information Officer for IT Planning, Architecture and E-Government Office of the Chief Information Officer Office of the Chief Information Officer 2/16/2010 2 TOPICS  Paperwork Reduction Act (PRA) of 1995 - Law  Paperwork Reduction Act - Overview  Information Collection Requests (ICRs)  Information

  10. Infrastructure Security EXCEPTIONAL SERVICE IN THE NATIONAL INTEREST

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    A New Age of Risk Assessment A probabilistic risk assessment (PRA), which is sometimes referred to as probabilistic safety assessment (PSA), is a systematic, logical analysis process. This powerful tool helps ensure the safe design and operation of complex engineered systems that have the potential for significant failure consequences. Sandia National Laboratories is a leader in the development and application of PRA methods in both civilian nuclear power and defense applications. In the past,

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

    SciTech Connect (OSTI)

    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.

  12. Systems Analysis Programs for Hands-On Integrated Reliability Evaluations.

    Energy Science and Technology Software Center (OSTI)

    2014-08-01

    Version 00 The U.S. Nuclear Regulatory Commission (NRC) has developed a powerful personal computer (PC) software application for performing probabilistic risk assessments (PRAs), called Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Version 8. Using SAPHIRE 8 on a PC, an analyst can perform a PRA for any complex system, facility, or process. Regarding nuclear power plants, SAPHIRE can be used to model a plant's response to initiating events, quantify associated core damage frequencies,more » and identify important contributors to core damage (Level 1 PRA). It can also be used to evaluate containment failure and release models for severe accident conditions, given that core damage has occurred (Level 2 PRA). It can be used for a PRA assuming that the reactor is at full power, at low power, or at shutdown conditions. Furthermore, it can be used to analyze both internal and external initiating events, and it 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 for release consequences to both the public and the environment (Level 3 PRA). For all of these models, SAPHIRE can evaluate the uncertainty inherent in the probabilistic models. SAPHIRE has evolved with advances in computer technology.« less

  13. New Methods and Tools to Perform Safety Analysis within RISMC

    SciTech Connect (OSTI)

    Diego Mandelli; Curtis Smith; Cristian Rabiti; Andrea Alfonsi; Robert Kinoshita; Joshua Cogliati

    2013-11-01

    The Risk Informed Safety Margins Characterization (RISMC) Pathway uses a systematic approach developed to characterize and quantify safety margins of nuclear power plant structures, systems and components. What differentiates the RISMC approach from traditional probabilistic risk assessment (PRA) is the concept of safety margin. In PRA, a safety metric such as core damage frequency (CDF) is generally estimated using static fault-tree and event-tree models. However, it is not possible to estimate how close we are to physical safety limits (say peak clad temperature) for most accident sequences described in the PRA. In the RISMC approach, what we want to understand is not just the frequency of an event like core damage, but how close we are (or not) to this event and how we might increase our safety margin through margin management strategies in a Dynamic PRA (DPRA) fashion. This paper gives an overview of methods that are currently under development at the Idaho National Laboratory (INL) with the scope of advance the current state of the art of dynamic PRA.

  14. Peer Review of NRC Standardized Plant Analysis Risk Models

    SciTech Connect (OSTI)

    Anthony Koonce; James Knudsen; Robert Buell

    2011-03-01

    The Nuclear Regulatory Commission (NRC) Standardized Plant Analysis Risk (SPAR) Models underwent a Peer Review using ASME PRA standard (Addendum C) as endorsed by NRC in Regulatory Guide (RG) 1.200. The review was performed by a mix of industry probabilistic risk analysis (PRA) experts and NRC PRA experts. Representative SPAR models, one PWR and one BWR, were reviewed against Capability Category I of the ASME PRA standard. Capability Category I was selected as the basis for review due to the specific uses/applications of the SPAR models. The BWR SPAR model was reviewed against 331 ASME PRA Standard Supporting Requirements; however, based on the Capability Category I level of review and the absence of internal flooding and containment performance (LERF) logic only 216 requirements were determined to be applicable. Based on the review, the BWR SPAR model met 139 of the 216 supporting requirements. The review also generated 200 findings or suggestions. Of these 200 findings and suggestions 142 were findings and 58 were suggestions. The PWR SPAR model was also evaluated against the same 331 ASME PRA Standard Supporting Requirements. Of these requirements only 215 were deemed appropriate for the review (for the same reason as noted for the BWR). The PWR review determined that 125 of the 215 supporting requirements met Capability Category I or greater. The review identified 101 findings or suggestions (76 findings and 25 suggestions). These findings or suggestions were developed to identify areas where SPAR models could be enhanced. A process to prioritize and incorporate the findings/suggestions supporting requirements into the SPAR models is being developed. The prioritization process focuses on those findings that will enhance the accuracy, completeness and usability of the SPAR models.

  15. SAPHIRE 8 Volume 3 - Users' Guide

    SciTech Connect (OSTI)

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

    2011-03-01

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

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

    SciTech Connect (OSTI)

    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.

  17. Low Power and Shutdown Risk Assessment Benchmarking Study

    SciTech Connect (OSTI)

    J.Mitman, J. Julius, R. Berucio, M. Phillips, J. Grobbelaaar, D. Bley, R. Budniz

    2002-12-15

    (B204)Probabilistic risk assessment (PRA) insights are now used by the United States Nuclear Regulatory Commission (USNRC) to confirm the level of safety for plant operations and to justify changes in nuclear power plant operating requirements, both on an exception basis and as changeds to a plant's licensing basis. This report examines qualitative and quantitative risk assessments during shutdown plant states, providing feedback to utilities in the use of qualitative models for outage risk management, and also providing input to the development of the American Nuclear Society (ANS) Low Power and Shutdown PRA Standard.

  18. A review of the Crystal River Unit 3 Probabilistic Risk Assessment: Internal events, core damage frequency

    SciTech Connect (OSTI)

    Hanan, N.A.; Henley, D.R.

    1989-01-01

    A review of the Crystal River Unit 3 Probabilistic Risk Assessment (CR-3 PRA) was performed with the objective of evaluating the dominant accident sequences and major contributions to the core damage frequency from internally-generated initiators. This review included not only an assessment of the assumption and methods used in the CR-3 PRA, but also included a quantitative analysis of the accident initiators, and accident sequences resulting in core damage. The effects of data uncertainties on the core damage frequency were quantified and sensitivity analysis was also performed. 55 refs., 22 figs., 30 tabs.

  19. May 16, 2016- Predicting the Service Life of Geomembranes in Low-Level and Mixed-Waste Disposal Facilities: Findings from a Long-Term Study

    Broader source: Energy.gov [DOE]

    Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - May 16, 2016 - Predicting the Service Life of Geomembranes in Low-Level and Mixed-Waste Disposal Facilities: Findings from a Long-Term Study. Presented by Dr. Craig Benson (Dean of School of Engineering and Applied Science, and Janet Scott Hamilton and John Downman Hamilton Professor, Univ. of Virginia).

  20. SAPHIRE 8 Volume 7 - Data Loading

    SciTech Connect (OSTI)

    K. J. Kvarfordt; S. T. Wood; C. L. Smith; S. R. Prescott

    2011-03-01

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE Version 8 is funded by the U.S. Nuclear Regulatory Commission and developed by the Idaho National Laboratory. This report is intended to assist the user to enter PRA data into the SAPHIRE program using the built-in MAR-D ASCII-text file data transfer process. Towards this end, a small sample database is constructed and utilized for demonstration. Where applicable, the discussion includes how the data processes for loading the sample database relate to the actual processes used to load a larger PRA models. The procedures described herein were developed for use with SAPHIRE Version 8. The guidance specified in this document will allow a user to have sufficient knowledge to both understand the data format used by SAPHIRE and to carry out the transfer of data between different PRA projects.

  1. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Data Loading Manual

    SciTech Connect (OSTI)

    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 primarily funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory. This report is intended to assist the user to enter PRA data into the SAPHIRE program using the built-in MAR-D ASCII-text file data transfer process. Towards this end, a small sample database is constructed and utilized for demonstration. Where applicable, the discussion includes how the data processes for loading the sample database relate to the actual processes used to load a larger PRA models. The procedures described herein were developed for use with SAPHIRE Version 6.0 and Version 7.0. In general, the data transfer procedures for version 6 and 7 are the same, but where deviations exist, the differences are noted. The guidance specified in this document will allow a user to have sufficient knowledge to both understand the data format used by SAPHIRE and to carry out the transfer of data between different PRA projects.

  2. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Data Loading Manual

    SciTech Connect (OSTI)

    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 primarily funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory. This report is intended to assist the user to enter PRA data into the SAPHIRE program using the built-in MAR-D ASCII-text file data transfer process. Towards this end, a small sample database is constructed and utilized for demonstration. Where applicable, the discussion includes how the data processes for loading the sample database relate to the actual processes used to load a larger PRA models. The procedures described herein were developed for use with SAPHIRE Version 6.0 and Version 7.0. In general, the data transfer procedures for version 6 and 7 are the same, but where deviations exist, the differences are noted. The guidance specified in this document will allow a user to have sufficient knowledge to both understand the data format used by SAPHIRE and to carry out the transfer of data between different PRA projects.

  3. August 18, 2015 Webinar- Probabilistic Analysis of Inadvertent Intrusion and the International Atomic Energy Agency Human Intrusion in the Context of Disposal of Radioactive Waste (HIDRA) Project

    Broader source: Energy.gov [DOE]

    P&RA CoP Webinar - August 18, 2015 - Probabilistic Analysis of Inadvertent Intrusion and the International Atomic Energy Agency Human Intrusion in the Context of Disposal of Radioactive Waste (HIDRA) Project, by Dr. Paul Black (Neptune) and Mr. Roger Seitz (Savannah River National Laboratory), August 18, 2015, 1:30 – 3:00 pm Eastern Daylight Time.

  4. May 16, 2016 Webinar- Predicting the Service Life of Geomembranes in Low-Level and Mixed-Waste Disposal Facilities

    Office of Energy Efficiency and Renewable Energy (EERE)

    Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - May 16, 2016 - Predicting the Service Life of Geomembranes in Low-Level and Mixed-Waste Disposal Facilities: Findings from a Long-Term Study. Presented by Dr. Craig Benson (Dean of School of Engineering and Applied Science, and Janet Scott Hamilton and John Downman Hamilton Professor, Univ. of Virginia).

  5. Analyzing system safety in lithium-ion grid energy storage

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Rosewater, David; Williams, Adam

    2015-10-08

    As grid energy storage systems become more complex, it grows more di cult to design them for safe operation. This paper first reviews the properties of lithium-ion batteries that can produce hazards in grid scale systems. Then the conventional safety engineering technique Probabilistic Risk Assessment (PRA) is reviewed to identify its limitations in complex systems. To address this gap, new research is presented on the application of Systems-Theoretic Process Analysis (STPA) to a lithium-ion battery based grid energy storage system. STPA is anticipated to ll the gaps recognized in PRA for designing complex systems and hence be more e ectivemore » or less costly to use during safety engineering. It was observed that STPA is able to capture causal scenarios for accidents not identified using PRA. Additionally, STPA enabled a more rational assessment of uncertainty (all that is not known) thereby promoting a healthy skepticism of design assumptions. Lastly, we conclude that STPA may indeed be more cost effective than PRA for safety engineering in lithium-ion battery systems. However, further research is needed to determine if this approach actually reduces safety engineering costs in development, or improves industry safety standards.« less

  6. Analyzing system safety in lithium-ion grid energy storage

    SciTech Connect (OSTI)

    Rosewater, David; Williams, Adam

    2015-10-08

    As grid energy storage systems become more complex, it grows more di cult to design them for safe operation. This paper first reviews the properties of lithium-ion batteries that can produce hazards in grid scale systems. Then the conventional safety engineering technique Probabilistic Risk Assessment (PRA) is reviewed to identify its limitations in complex systems. To address this gap, new research is presented on the application of Systems-Theoretic Process Analysis (STPA) to a lithium-ion battery based grid energy storage system. STPA is anticipated to ll the gaps recognized in PRA for designing complex systems and hence be more e ective or less costly to use during safety engineering. It was observed that STPA is able to capture causal scenarios for accidents not identified using PRA. Additionally, STPA enabled a more rational assessment of uncertainty (all that is not known) thereby promoting a healthy skepticism of design assumptions. Lastly, we conclude that STPA may indeed be more cost effective than PRA for safety engineering in lithium-ion battery systems. However, further research is needed to determine if this approach actually reduces safety engineering costs in development, or improves industry safety standards.

  7. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

    SciTech Connect (OSTI)

    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 systems 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 general

  8. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) Code Reference Manual

    SciTech Connect (OSTI)

    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 systems 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 general

  9. Licensing topical report: application of probabilistic risk assessment in the selection of design basis accidents. [HTGR

    SciTech Connect (OSTI)

    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 (HTGRs).

  10. System 80+ D-RAP, a communication tool

    SciTech Connect (OSTI)

    Siegmann, E.R.; Mody, A.A.

    1994-12-31

    The purpose of {open_quotes}RAP{close_quotes} music is to communicate, and the purpose of D-RAP is to foster communication between the probabilistic risk assessment (PRA) group, designers, and the future combined operating license (COL) applicant. This is to ensure that the design is self-consistent and integrated with the procurement process. The designer reliability assurance program (D-RAP) is the first part of the RAP. The goals of the D-RAP are to have risk-significant systems, structures, and components (SSCs) identified and considered in the detail design and procurement phases and to maintain consistency between PRA and design. Plant safety is maintained throughout the design phase, and pertinent information is passed on to the COL applicant. The operations RAP (O-RAP) covers the plant operation and maintenance.

  11. Need for a probabilistic risk assessment of the oil tanker industry and a qualitative assessment of oil tanker groundings. Master`s thesis

    SciTech Connect (OSTI)

    Amrozowiez, M.D.

    1996-06-01

    The culture, design, and operation of the maritime industry all contribute to create an error-inducing system. A probabilistic risk assessment (PRA) provides a formal process of determining the full range of possible adverse occurrences, probabilities, and expected costs for any undesirable event. A PRA can identify those areas that offer the greatest risk-reducing potential. Once the components with the greatest risk-reducing potential are identified, appropriate technology and management schemes can properly influence risk reduction. While human error is attributed to 80 percent of the marine accidents, a closer look reveals that many accidents attributed to human error are system errors. An application of a qualitative risk assessment is done for tanker groundings. A fault tree is developed to describe the top event of a tanker grounding. A number of well-known groundings are analyzed to test the utility of the grounding fault tree.

  12. SAPHIRE 8 Software Quality Assurance Plan

    SciTech Connect (OSTI)

    Curtis Smith

    2010-02-01

    This Quality Assurance (QA) Plan documents the QA activities that will be managed by the INL related to JCN N6423. The NRC developed the SAPHIRE computer code for performing probabilistic risk assessments (PRAs) using a personal computer (PC) at the Idaho National Laboratory (INL) under Job Code Number (JCN) L1429. SAPHIRE started out as a feasibility study for a PRA code to be run on a desktop personal PC and evolved through several phases into a state-of-the-art PRA code. The developmental activity of SAPHIRE was the result of two concurrent important events: The tremendous expansion of PC software and hardware capability of the 90s and the onset of a risk-informed regulation era.

  13. A Program for Risk Assessment Associated with IGSCC of BWR Vessel Internals

    SciTech Connect (OSTI)

    A. G. Ware; D. K. Morton; J. D. Page; M. E. Nitzel; S. A. Eide; T. -Y. Chang

    1999-08-01

    A program is being carried out for the US Nuclear Regulatory Commission (NRC) by the Idaho National Engineering and Environmental Laboratory (INEEL), to conduct an independent risk assessment of the consequences of failures initiated by intergranular stress corrosion cracking (IGSCC) of the reactor vessel internals of boiling water reactor (BWR) plants. The overall project objective is to assess the potential consequences and risks associated with the failure of IGSCC-susceptible BWR vessel internals, both singly and in combination with the failures of others, with specific consideration given to potential cascading and common mode effects on system performance. This paper presents a description of the overall program that is underway to modify an existing probabilistic risk assessment (PRA) of the BWR/4 plant to include IGSCC-initiated failures, subsequently to complete a quantitative PRA.

  14. Methodology Development for Passive Component Reliability Modeling in a Multi-Physics Simulation Environment

    SciTech Connect (OSTI)

    Aldemir, Tunc; Denning, Richard; Catalyurek, Umit; Unwin, Stephen

    2015-01-23

    Reduction in safety margin can be expected as passive structures and components undergo degradation with time. Limitations in the traditional probabilistic risk assessment (PRA) methodology constrain its value as an effective tool to address the impact of aging effects on risk and for quantifying the impact of aging management strategies in maintaining safety margins. A methodology has been developed to address multiple aging mechanisms involving large numbers of components (with possibly statistically dependent failures) within the PRA framework in a computationally feasible manner when the sequencing of events is conditioned on the physical conditions predicted in a simulation environment, such as the New Generation System Code (NGSC) concept. Both epistemic and aleatory uncertainties can be accounted for within the same phenomenological framework and maintenance can be accounted for in a coherent fashion. The framework accommodates the prospective impacts of various intervention strategies such as testing, maintenance, and refurbishment. The methodology is illustrated with several examples.

  15. Probabilistic risk analysis toward cost-effective 3S (safety, safeguards, security) implementation

    SciTech Connect (OSTI)

    Suzuki, Mitsutoshi; Mochiji, Toshiro

    2014-09-30

    Probabilistic Risk Analysis (PRA) has been introduced for several decades in safety and nuclear advanced countries have already used this methodology in their own regulatory systems. However, PRA has not been developed in safeguards and security so far because of inherent difficulties in intentional and malicious acts. In this paper, probabilistic proliferation and risk analysis based on random process is applied to hypothetical reprocessing process and physical protection system in nuclear reactor with the Markov model that was originally developed by the Proliferation Resistance and Physical Protection Working Group (PRPPWG) in Generation IV International Framework (GIF). Through the challenge to quantify the security risk with a frequency in this model, integrated risk notion among 3S to pursue the cost-effective installation of those countermeasures is discussed in a heroic manner.

  16. Maintenance personnel performance simulation (MAPPS) model: overview and evaluation efforts

    SciTech Connect (OSTI)

    Knee, H.E.; Haas, P.M.; Siegel, A.I.; Bartter, W.D.; Wolf, J.J.; Ryan, T.G.

    1984-01-01

    The development of the MAPPS model has been completed and the model is currently undergoing evaluation. These efforts are addressing a number of identified issues concerning practicality, acceptability, usefulness, and validity. Preliminary analysis of the evaluation data that has been collected indicates that MAPPS will provide comprehensive and reliable data for PRA purposes and for a number of other applications. The MAPPS computer simulation model provides the user with a sophisticated tool for gaining insights into tasks performed by NPP maintenance personnel. Its wide variety of input parameters and output data makes it extremely flexible for application to a number of diverse applications. With the demonstration of favorable model evaluation results, the MAPPS model will represent a valuable source of NPP maintainer reliability data and provide PRA studies with a source of data on maintainers that has previously not existed.

  17. Information Collection Management | Department of Energy

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    Information Collection Management Information Collection Management Forms44.jpg The Paperwork Reduction Act (PRA) of 1995 requires federal agencies and government-owned, contractor-operated facilities to obtain approval from the Office of Management and Budget (OMB) before collecting information from the general public, which includes contractors. This ensures each organization will maximize the utility of information created, collected, maintained, and used. The coordination with OMB also

  18. A probabilistic risk assessment of the LLNL Plutonium facility`s evaluation basis fire operational accident

    SciTech Connect (OSTI)

    Brumburgh, G.

    1994-08-31

    The Lawrence Livermore National Laboratory (LLNL) Plutonium Facility conducts numerous involving plutonium to include device fabrication, development of fabrication techniques, metallurgy research, and laser isotope separation. A Safety Analysis Report (SAR) for the building 332 Plutonium Facility was completed rational safety and acceptable risk to employees, the public, government property, and the environment. This paper outlines the PRA analysis of the Evaluation Basis Fire (EDF) operational accident. The EBF postulates the worst-case programmatic impact event for the Plutonium Facility.

  19. Applicability of Loss of Offsite Power (LOSP) Events in NUREG/CR-6890 for Entergy Nuclear South (ENS) Plants LOSP Calculations

    SciTech Connect (OSTI)

    Li, Yunlong; Yilmaz, Fatma; Bedell, Loys

    2006-07-01

    Significant differences have been identified in loss of offsite power (LOSP or LOOP) event description, category, duration, and applicability between the LOSP events used in NUREG/CR-6890 and ENS'LOSP packages, which were based on EPRI LOSP reports with plant-specific applicability analysis. Thus it is appropriate to reconcile the LOSP data listed in the subject NUREG and EPRI reports. A cross comparison showed that 62 LOSP events in NUREG/CR-6890 were not included in the EPRI reports while 4 events in EPRI reports were missing in the NUREG. Among the 62 events missing in EPRI reports, the majority were applicable to shutdown conditions, which could be classified as category IV events in EPRI reports if included. Detailed reviews of LERs concluded that some events did not result in total loss of offsite power. Some LOSP events were caused by subsequent component failures after a turbine/plant trip, which have been modeled specifically in most ENS plant PRA models. Moreover, ENS has modeled (or is going to model) the partial loss of offsite power events with partial LOSP initiating events. While the direct use of NUREG/CR-6890 results in SPAR models may be appropriate, its direct use in ENS' plant PRA models may not be appropriate because of modeling details in ENS' plant-specific PRA models. Therefore, this paper lists all the differences between the data in NUREG/CR-6890 and EPRI reports and evaluates the applicability of the LOSP events to ENS plant-specific PRA models. The refined LOSP data will characterize the LOSP risk in a more realistic fashion. (authors)

  20. Dynamic Event Tree Analysis Through RAVEN

    SciTech Connect (OSTI)

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

    2013-09-01

    Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics is not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (D-PRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed in a high modular and pluggable way in order to enable easy integration of different programming languages (i.e., C++, Python) and coupling with other application including the ones based on the MOOSE framework, developed by INL as well. RAVEN performs two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, RAVEN also models stochastic events, such as components failures, and performs uncertainty quantification. Such stochastic modeling is employed by using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This paper focuses on the first task and shows how it is possible to perform the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, the Dynamic PRA analysis, using Dynamic Event Tree, of a simplified pressurized water reactor for a Station Black-Out scenario is presented.

  1. fe0025387-Petrotechnical-Resources | netl.doe.gov

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Support for Methane Hydrate Research on the Alaska North Slope Last Reviewed 1/15/2016 DE-FE0025387 Goal The objective of this contract between the Department of Energy (DOE) and Petrotechnical Resources of Alaska (PRA) is to provide specific planning, analytical, arctic engineering, and environmental services associated with the potential drilling of methane hydrate stratigraphic test well(s) and a long-term methane hydrate production test well on the North Slope of Alaska. Performer

  2. ORISE: Advanced Radiation Medicine | REAC/TS Continuing Medical Education

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Course Advanced Radiation Medicine Dates Scheduled Register Online April 24-28, 2017 August 14-18, 2017 Fee: $275 Maximum enrollment: 28 30 hours AMA PRA Category 1 Credits(tm) This 4½-day course includes more advanced information for medical practitioners. This program is academically more rigorous than the REM course and is primarily for Physicians, Physician Assistants, Nurse Practitioners, and Nurses desiring an advanced level of information on the diagnosis and management of ionizing

  3. Turning points in reactor design

    SciTech Connect (OSTI)

    Beckjord, E.S.

    1995-09-01

    This article provides some historical aspects on nuclear reactor design, beginning with PWR development for Naval Propulsion and the first commercial application at Yankee Rowe. Five turning points in reactor design and some safety problems associated with them are reviewed: (1) stability of Dresden-1, (2) ECCS, (3) PRA, (4) TMI-2, and (5) advanced passive LWR designs. While the emphasis is on the thermal-hydraulic aspects, the discussion is also about reactor systems.

  4. Robustness of RISMC Insights under Alternative Aleatory/Epistemic Uncertainty Classifications: Draft Report under the Risk-Informed Safety Margin Characterization (RISMC) Pathway of the DOE Light Water Reactor Sustainability Program

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Eslinger, Paul W.; Johnson, Kenneth I.

    2012-09-20

    The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy (DOE) Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A technical challenge at the core of this effort is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, would be founded on probabilistic characterizations of uncertainty in SSC performance. In the context of probabilistic risk assessment (PRA) technology, there has arisen a general consensus about the distinctive roles of two types of uncertainty: aleatory and epistemic, where the former represents irreducible, random variability inherent in a system, whereas the latter represents a state of knowledge uncertainty on the part of the analyst about the system which is, in principle, reducible through further research. While there is often some ambiguity about how any one contributing uncertainty in an analysis should be classified, there has nevertheless emerged a broad consensus on the meanings of these uncertainty types in the PRA setting. However, while RISMC methodology shares some features with conventional PRA, it will nevertheless be a distinctive methodology set. Therefore, the paradigms for classification of uncertainty in the PRA setting may not fully port to the RISMC environment. Yet the notion of risk-informed margin is based on the characterization of uncertainty, and it is therefore critical to establish a common understanding of uncertainty in the RISMC setting.

  5. DOE F 4200.33.cdr

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    3 (07-05) U.S. Department of Energy Note: ** We Hereby Certify That Funds Cited Are Proper For This Procurement And In Compliance With Applicable Appropriations Acts and Fiscal Law. Printed with soy ink on recycled paper Procurement Request-Authorization 1. Awarding Office Fund Year Alottee Reporting Entity SGL Object Class Program Project WFO Local Use 26. Dollar Amount 27. Program Budget Official's Signature** 3. PRA Number Formerly PR-799A (Previous editions are obsolete) 2. Initiating Office

  6. DOE F 4200.34.cdr

    Office of Energy Efficiency and Renewable Energy (EERE) Indexed Site

    4 (07-05) U.S. Department of Energy ** Note: We Hereby Certify That Funds Cited Are Proper For This Procurement And In Compliance With Applicable Appropriations Acts and Fiscal Law. Printed with soy ink on recycled paper Procurement Request-Authorization Funding Data Continuation Sheet 1. PRA Number 2. Change/Correction in Process Yes No 25. ACCOUNTING AND APPROPRIATION DATA Fund Year Alottee Reporting Entity SGL Object Class Program Project WFO Local Use 26. Dollar Amount 27. Program Budget

  7. Severe accident progression perspectives based on IPE results

    SciTech Connect (OSTI)

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

    1996-08-01

    Accident progression perspectives were gathered from the level 2 PRA analyses (the analysis of the accident after core damage has occurred involving the containment performance and the radionuclide release from the containment) described in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were obtained. Complete results are discussed in NUREG-1560 and summarized here.

  8. Containment performance perspectives based on IPE results

    SciTech Connect (OSTI)

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

    1996-12-31

    Perspectives on Containment Performance were obtained from the accident progression analyses, i.e. level 2 PRA analyses, found in the IPE submittals. Insights related to the containment failure modes, the releases associated with those failure modes, and the factors responsible for the types of containment failures and release sizes reported were gathered. The results summarized here are discussed in detail in volumes 1 and 2 of NUREG 1560. 3 refs., 4 figs.

  9. A REVIEW OF SOFTWARE-INDUCED FAILURE EXPERIENCE.

    SciTech Connect (OSTI)

    CHU, T.L.; MARTINEZ-GURIDI, G.; YUE, M.; LEHNER, J.

    2006-09-01

    We present a review of software-induced failures in commercial nuclear power plants (NPPs) and in several non-nuclear industries. We discuss the approach used for connecting operational events related to these failures and the insights gained from this review. In particular, we elaborate on insights that can be used to model this kind of failure in a probabilistic risk assessment (PRA) model. We present the conclusions reached in these areas.

  10. A Research Roadmap for Computation-Based Human Reliability Analysis

    SciTech Connect (OSTI)

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

    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.

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

    SciTech Connect (OSTI)

    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.

  12. Capabilities needed for the next generation of thermo-hydraulic codes for use in real time applications

    SciTech Connect (OSTI)

    Arndt, S.A.

    1997-07-01

    The real-time reactor simulation field is currently at a crossroads in terms of the capability to perform real-time analysis using the most sophisticated computer codes. Current generation safety analysis codes are being modified to replace simplified codes that were specifically designed to meet the competing requirement for real-time applications. The next generation of thermo-hydraulic codes will need to have included in their specifications the specific requirement for use in a real-time environment. Use of the codes in real-time applications imposes much stricter requirements on robustness, reliability and repeatability than do design and analysis applications. In addition, the need for code use by a variety of users is a critical issue for real-time users, trainers and emergency planners who currently use real-time simulation, and PRA practitioners who will increasingly use real-time simulation for evaluating PRA success criteria in near real-time to validate PRA results for specific configurations and plant system unavailabilities.

  13. WHEN MODEL MEETS REALITY – A REVIEW OF SPAR LEVEL 2 MODEL AGAINST FUKUSHIMA ACCIDENT

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01

    The Standardized Plant Analysis Risk (SPAR) models are a set of probabilistic risk assessment (PRA) models used by the Nuclear Regulatory Commission (NRC) to evaluate the risk of operations at U.S. nuclear power plants and provide inputs to risk informed regulatory process. A small number of SPAR Level 2 models have been developed mostly for feasibility study purpose. They extend the Level 1 models to include containment systems, group plant damage states, and model containment phenomenology and accident progression in containment event trees. A severe earthquake and tsunami hit the eastern coast of Japan in March 2011 and caused significant damages on the reactors in Fukushima Daiichi site. Station blackout (SBO), core damage, containment damage, hydrogen explosion, and intensive radioactivity release, which have been previous analyzed and assumed as postulated accident progression in PRA models, now occurred with various degrees in the multi-units Fukushima Daiichi site. This paper reviews and compares a typical BWR SPAR Level 2 model with the “real” accident progressions and sequences occurred in Fukushima Daiichi Units 1, 2, and 3. It shows that the SPAR Level 2 model is a robust PRA model that could very reasonably describe the accident progression for a real and complicated nuclear accident in the world. On the other hand, the comparison shows that the SPAR model could be enhanced by incorporating some accident characteristics for better representation of severe accident progression.

  14. A technique for human error analysis (ATHEANA)

    SciTech Connect (OSTI)

    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.

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

    SciTech Connect (OSTI)

    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.

  16. Auxiliary feedwater system risk-based inspection guide for the South Texas Project nuclear power plant

    SciTech Connect (OSTI)

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

    1993-12-01

    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. South Texas Project 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 South Texas Project plant.

  17. Auxiliary feedwater system risk-based inspection guide for the McGuire nuclear power plant

    SciTech Connect (OSTI)

    Bumgardner, J.D.; Lloyd, R.C.; Moffitt, N.E.; Gore, B.F.; Vo, T.V.

    1994-05-01

    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. McGuire 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 McGuire plant.

  18. Auxiliary feedwater system risk-based inspection guide for the Point Beach nuclear power plant

    SciTech Connect (OSTI)

    Lloyd, R C; Moffitt, N E; Gore, B F; Vo, T V; Vehec, T 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. Point Beach 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 Point Beach plant.

  19. Auxiliary feedwater system risk-based inspection guide for the J. M. Farley Nuclear Power Plant

    SciTech Connect (OSTI)

    Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G. )

    1990-10-01

    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 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. J. M. Farley was selected as the second 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 at the J. M. Farley plant. 23 refs., 1 fig., 1 tab.

  20. Auxiliary feedwater system risk-based inspection guide for the H. B. Robinson nuclear power plant

    SciTech Connect (OSTI)

    Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V.; Garner, L.W.

    1993-08-01

    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. H. B. Robinson 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 H. B. Robinson plant.

  1. Auxiliary feedwater system risk-based inspection guide for the Byron and Braidwood nuclear power plants

    SciTech Connect (OSTI)

    Moffitt, N.E.; Gore, B.F.: Vo, T.V. )

    1991-07-01

    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. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants 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 Byron/Braidwood plants. 23 refs., 1 fig., 1 tab.

  2. Methodology for the Incorporation of Passive Component Aging Modeling into the RAVEN/ RELAP-7 Environment

    SciTech Connect (OSTI)

    Mandelli, Diego; Rabiti, Cristian; Cogliati, Joshua; Alfonsi, Andrea; Askin Guler; Tunc Aldemir

    2014-11-01

    Passive system, structure and components (SSCs) will degrade over their operation life and this degradation may cause to reduction in the safety margins of a nuclear power plant. In traditional probabilistic risk assessment (PRA) using the event-tree/fault-tree methodology, passive SSC failure rates are generally based on generic plant failure data and the true state of a specific plant is not reflected realistically. To address aging effects of passive SSCs in the traditional PRA methodology [1] does consider physics based models that account for the operating conditions in the plant, however, [1] does not include effects of surveillance/inspection. This paper represents an overall methodology for the incorporation of aging modeling of passive components into the RAVEN/RELAP-7 environment which provides a framework for performing dynamic PRA. Dynamic PRA allows consideration of both epistemic and aleatory uncertainties (including those associated with maintenance activities) in a consistent phenomenological and probabilistic framework and is often needed when there is complex process/hardware/software/firmware/ human interaction [2]. Dynamic PRA has gained attention recently due to difficulties in the traditional PRA modeling of aging effects of passive components using physics based models and also in the modeling of digital instrumentation and control systems. RAVEN (Reactor Analysis and Virtual control Environment) [3] is a software package under development at the Idaho National Laboratory (INL) as an online control logic driver and post-processing tool. It is coupled to the plant transient code RELAP-7 (Reactor Excursion and Leak Analysis Program) also currently under development at INL [3], as well as RELAP 5 [4]. The overall methodology aims to: • Address multiple aging mechanisms involving large number of components in a computational feasible manner where sequencing of events is conditioned on the physical conditions predicted in a simulation

  3. Risk-Informed Safety Margin Characterization Methods Development Work

    SciTech Connect (OSTI)

    Smith, Curtis L; Ma, Zhegang; Tom Riley; Mandelli, Diego; Nielsen, Joseph W; Alfonsi, Andrea; Rabiti, Cristian

    2014-09-01

    This report summarizes the research activity developed during the Fiscal year 2014 within the Risk Informed Safety Margin and Characterization (RISMC) pathway within the Light Water Reactor Sustainability (LWRS) campaign. This research activity is complementary to the one presented in the INL/EXT-??? report which shows advances Probabilistic Risk Assessment Analysis using RAVEN and RELAP-7 in conjunction to novel flooding simulation tools. Here we present several analyses that prove the values of the RISMC approach in order to assess risk associated to nuclear power plants (NPPs). We focus on simulation based PRA which, in contrast to classical PRA, heavily employs system simulator codes. Firstly we compare, these two types of analyses, classical and RISMC, for a Boiling water reactor (BWR) station black out (SBO) initiating event. Secondly we present an extended BWR SBO analysis using RAVEN and RELAP-5 which address the comments and suggestions received about he original analysis presented in INL/EXT-???. This time we focus more on the stochastic analysis such probability of core damage and on the determination of the most risk-relevant factors. We also show some preliminary results regarding the comparison between RELAP5-3D and the new code RELAP-7 for a simplified Pressurized Water Reactors system. Lastly we present some conceptual ideas regarding the possibility to extended the RISMC capabilities from an off-line tool (i.e., as PRA analysis tool) to an online-tool. In this new configuration, RISMC capabilities can be used to assist and inform reactor operator during real accident scenarios.

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

    SciTech Connect (OSTI)

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

    1995-03-01

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

  5. Review of Quantitative Software Reliability Methods

    SciTech Connect (OSTI)

    Chu, T.L.; Yue, M.; Martinez-Guridi, M.; Lehner, J.

    2010-09-17

    The current U.S. Nuclear Regulatory Commission (NRC) licensing process for digital systems rests on deterministic engineering criteria. In its 1995 probabilistic risk assessment (PRA) policy statement, the Commission encouraged the use of PRA technology in all regulatory matters to the extent supported by the state-of-the-art in PRA methods and data. Although many activities have been completed in the area of risk-informed regulation, the risk-informed analysis process for digital systems has not yet been satisfactorily developed. Since digital instrumentation and control (I&C) systems are expected to play an increasingly important role in nuclear power plant (NPP) safety, the NRC established a digital system research plan that defines a coherent set of research programs to support its regulatory needs. One of the research programs included in the NRC's digital system research plan addresses risk assessment methods and data for digital systems. Digital I&C systems have some unique characteristics, such as using software, and may have different failure causes and/or modes than analog I&C systems; hence, their incorporation into NPP PRAs entails special challenges. The objective of the NRC's digital system risk research is to identify and develop methods, analytical tools, and regulatory guidance for (1) including models of digital systems into NPP PRAs, and (2) using information on the risks of digital systems to support the NRC's risk-informed licensing and oversight activities. For several years, Brookhaven National Laboratory (BNL) has worked on NRC projects to investigate methods and tools for the probabilistic modeling of digital systems, as documented mainly in NUREG/CR-6962 and NUREG/CR-6997. However, the scope of this research principally focused on hardware failures, with limited reviews of software failure experience and software reliability methods. NRC also sponsored research at the Ohio State University investigating the modeling of digital systems

  6. Review of the Diablo Canyon probabilistic risk assessment

    SciTech Connect (OSTI)

    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.

  7. ORISE: Radiation Emergency Medicine | REAC/TS Continuing Medical Education

    Broader source: All U.S. Department of Energy (DOE) Office Webpages (Extended Search)

    Course Radiation Emergency Medicine Dates Scheduled Register Online October 11-14, 2016 February 7-10, 2017 February 28 - March 3, 2017 April 18-21, 2017 June 13-16, 2017 August 8-11, 2017 Fee: $200 Maximum enrollment: 22 24.5 hours AMA PRA Category 1 Credits(tm) This 3½-day course is intended for Physicians, Physician Assistants, Nurse Practitioners, Nurses and other healthcare providers. First responders, emergency management, and public health professionals may find the course

  8. Seismic fragility analysis of structural components for HFBR facilities

    SciTech Connect (OSTI)

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

    1992-01-01

    The paper presents a summary of recently completed seismic fragility analyses of the HFBR facilities. Based on a detailed review of past PRA studies, various refinements were made regarding the strength and ductility evaluation of structural components. Available laboratory test data were analysed to evaluate the formulations used to predict the ultimate strength and deformation capacities of steel, reinforced concrete and masonry structures. The biasness and uncertainties were evaluated within the framework of the fragility evaluation methods widely accepted in the nuclear industry. A few examples of fragility calculations are also included to illustrate the use of the presented formulations.

  9. Seismic fragility analysis of structural components for HFBR facilities

    SciTech Connect (OSTI)

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

    1992-04-01

    The paper presents a summary of recently completed seismic fragility analyses of the HFBR facilities. Based on a detailed review of past PRA studies, various refinements were made regarding the strength and ductility evaluation of structural components. Available laboratory test data were analysed to evaluate the formulations used to predict the ultimate strength and deformation capacities of steel, reinforced concrete and masonry structures. The biasness and uncertainties were evaluated within the framework of the fragility evaluation methods widely accepted in the nuclear industry. A few examples of fragility calculations are also included to illustrate the use of the presented formulations.

  10. Adaptive Sampling using Support Vector Machines

    SciTech Connect (OSTI)

    D. Mandelli; C. Smith

    2012-11-01

    Reliability/safety analysis of stochastic dynamic systems (e.g., nuclear power plants, airplanes, chemical plants) is currently performed through a combination of Event-Tress and Fault-Trees. However, these conventional methods suffer from certain drawbacks: Timing of events is not explicitly modeled Ordering of events is preset by the analyst The modeling of complex accident scenarios is driven by expert-judgment For these reasons, there is currently an increasing interest into the development of dynamic PRA methodologies since they can be used to address the deficiencies of conventional methods listed above.

  11. Framework Development Supporting the Safety Portal

    SciTech Connect (OSTI)

    Prescott, Steven Ralph; Kvarfordt, Kellie Jean; Vang, Leng; Smith, Curtis Lee

    2015-07-01

    In a collaborating scientific research arena it is important to have an environment where analysts have access to a shared repository of information, documents, and software tools, and be able to accurately maintain and track historical changes in models. The new Safety Portal cloud-based environment will be accessible remotely from anywhere regardless of computing platforms given that the platform has available Internet access and proper browser capabilities. Information stored at this environment would be restricted based on user assigned credentials. This report discusses current development of a cloud-based web portal for PRA tools.

  12. Cloud-based Architecture Capabilities Summary Report

    SciTech Connect (OSTI)

    Vang, Leng; Prescott, Steven R; Smith, Curtis

    2014-09-01

    In collaborating scientific research arena it is important to have an environment where analysts have access to a shared of information documents, software tools and be able to accurately maintain and track historical changes in models. A new cloud-based environment would be accessible remotely from anywhere regardless of computing platforms given that the platform has available of Internet access and proper browser capabilities. Information stored at this environment would be restricted based on user assigned credentials. This report reviews development of a Cloud-based Architecture Capabilities (CAC) as a web portal for PRA tools.

  13. Optimization Method to Branch and Bound Large SBO State Spaces Under Dynamic Probabilistic Risk Assessment via use of LENDIT Scales and S2R2 Sets

    SciTech Connect (OSTI)

    Joseph W. Nielsen; Akira Tokurio; Robert Hiromoto; Jivan Khatry

    2014-06-01

    Traditional Probabilistic Risk Assessment (PRA) methods have been developed and are quite effective in evaluating risk associated with complex systems, but lack the capability to evaluate complex dynamic systems. These time and energy scales associated with the transient may vary as a function of transition time to a different physical state. Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems, while complete, results in issues associated with combinatorial explosion. In order to address the combinatorial complexity arising from the number of possible state configurations and discretization of transition times, a characteristic scaling metric (LENDIT length, energy, number, distribution, information and time) is proposed as a means to describe systems uniformly and thus provide means to describe relational constraints expected in the dynamics of a complex (coupled) systems. Thus when LENDIT is used to characterize four sets state, system, resource and response (S2R2) describing reactor operations (normal and off-normal), LENDIT and S2R2 in combination have the potential to branch and bound the state space investigated by DPRA. In this paper we introduce the concept of LENDIT scales and S2R2 sets applied to a branch-and-bound algorithm and apply the methods to a station black out transient (SBO).

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

    SciTech Connect (OSTI)

    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.

  15. Quantitative risk of oil tanker groundings. Master`s thesis

    SciTech Connect (OSTI)

    Amrozowicz, M.D.

    1996-06-01

    The culture, design, and operation of the maritime industry all contribute to create an error-inducing system. As oil tankers have become larger, the tolerance for error has decreased as the consequences have increased. Tankers are the largest contributor by vessel type to worldwide oil spill volume. Human error has consistently been attributed to 80 percent of the marine accidents. A closer look reveals that many accidents attributed to human error are system errors. In fact, the term human error is unwarranted in many high-risk accidents and its use is a perforation of the context. The maritime industry has been identified as a high risk operation, requiring an active risk management program. A probabilistic risk assessment (PRA) provides a formal process of determining the full range of possible adverse occurrences, probabilities, and expected costs for any undesirable event. A PRA can identify those areas that offer the greatest risk-reducing potential. This thesis focuses on the first level of a proposed three-level risk model to determine the probability of a tanker grounding. The approach utilizes fault trees and event trees and incorporates The Human Error Rate Prediction data to quantify individual errors. The result allows the identification of high-leverage factors in order to determine the most effective and efficient use of resources to reduce the probability of grounding; showing that the development of the Electronic Chart Display and Information System incorporated with the International Safety Management Code can significantly reduce the probability of grounding.

  16. Dynamical systems probabilistic risk assessment.

    SciTech Connect (OSTI)

    Denman, Matthew R.; Ames, Arlo Leroy

    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.

  17. Integrated Risk Assessment for the LaSalle Unit 2 Nuclear Power Plant, Phenomenology and Risk Uncertainty Evaluation Program (PRUEP), MELCOR code calculations. Volume 3

    SciTech Connect (OSTI)

    Shaffer, C.J. [Science and Engineering Associates, Albuquerque, NM (United States); Miller, L.A.; Payne, A.C. Jr.

    1992-10-01

    A Level III Probabilistic Risk Assessment (PRA) has been performed for LaSalle Unit 2 under the Risk Methods Integration and Evaluation Program (RMIEP) and the Phenomenology and Risk Uncertainty Evaluation Program (PRUEP). This report documents the phenomenological calculations and sources of. uncertainty in the calculations performed with HELCOR in support of the Level II portion of the PRA. These calculations are an integral part of the Level II analysis since they provide quantitative input to the Accident Progression Event Tree (APET) and Source Term Model (LASSOR). However, the uncertainty associated with the code results must be considered in the use of the results. The MELCOR calculations performed include four integrated calculations: (1) a high-pressure short-term station blackout, (2) a low-pressure short-term station blackout, (3) an intermediate-term station blackout, and (4) a long-term station blackout. Several sensitivity studies investigating the effect of variations in containment failure size and location, as well as hydrogen ignition concentration are also documented.

  18. Air ingression calculations for selected plant transients using MELCOR

    SciTech Connect (OSTI)

    Kmetyk, L.N.

    1994-01-01

    Two sets of MELCOR calculations have been completed studying the effects of air ingression on the consequences of various severe accident scenarios. One set of calculations analyzed a station blackout with surge line failure prior to vessel breach, starting from nominal operating conditions; the other set of calculations analyzed a station blackout occurring during shutdown (refueling) conditions. Both sets of analyses were for the Surry plant, a three-loop Westinghouse PWR. For both accident scenarios, a basecase calculation was done, and then repeated with air ingression from containment into the core region following core degradation and vessel failure. In addition to the two sets of analyses done for this program, a similar air-ingression sensitivity study was done as part of a low-power/shutdown PRA, with results summarized here; that PRA study also analyzed a station blackout occurring during shutdown (refueling) conditions, but for the Grand Gulf plant, a BWR/6 with Mark III containment. These studies help quantify the amount of air that would have to enter the core region to have a significant impact on the severe accident scenario, and demonstrate that one effect, of air ingression is substantial enhancement of ruthenium release. These calculations also show that, while the core clad temperatures rise more quickly due to oxidation with air rather than steam, the core also degrades and relocates more quickly, so that no sustained, enhanced core heatup is predicted to occur with air ingression.

  19. Analysis of Kuosheng Station Blackout Accident Using MELCOR 1.8.4

    SciTech Connect (OSTI)

    Wang, S.-J.; Chien, C.-S.; Wang, T.-C.; Chiang, K.-S

    2000-11-15

    The MELCOR code, developed by Sandia National Laboratories, is a fully integrated, relatively fast-running code that models the progression of severe accidents in commercial light water nuclear power plants (NPPs).A specific station blackout (SBO) accident for Kuosheng (BWR-6) NPP is simulated using the MELCOR 1.8.4 code. The MELCOR input deck for Kuosheng NPP is established based on Kuosheng NPP design data and the MELCOR users' guides. The initial steady-state conditions are generated with a developed self-initialization algorithm. The main severe accident phenomena and the fission product release fractions associated with the SBO accident were simulated. The predicted results are plausible and as expected in light of current understanding of severe accident phenomena. The uncertainty of this analysis is briefly discussed. The important features of the MELCOR 1.8.4 are described. The estimated results provide useful information for the probabilistic risk assessment (PRA) of Kuosheng NPP. This tool will be applied to the PRA, the severe accident analysis, and the severe accident management study of Kuosheng NPP in the near future.

  20. Integrated Risk Assessment for the LaSalle Unit 2 Nuclear Power Plant, Phenomenology and Risk Uncertainty Evaluation Program (PRUEP), MELCOR code calculations

    SciTech Connect (OSTI)

    Shaffer, C.J. (Science and Engineering Associates, Albuquerque, NM (United States)); Miller, L.A.; Payne, A.C. Jr.

    1992-10-01

    A Level III Probabilistic Risk Assessment (PRA) has been performed for LaSalle Unit 2 under the Risk Methods Integration and Evaluation Program (RMIEP) and the Phenomenology and Risk Uncertainty Evaluation Program (PRUEP). This report documents the phenomenological calculations and sources of. uncertainty in the calculations performed with HELCOR in support of the Level II portion of the PRA. These calculations are an integral part of the Level II analysis since they provide quantitative input to the Accident Progression Event Tree (APET) and Source Term Model (LASSOR). However, the uncertainty associated with the code results must be considered in the use of the results. The MELCOR calculations performed include four integrated calculations: (1) a high-pressure short-term station blackout, (2) a low-pressure short-term station blackout, (3) an intermediate-term station blackout, and (4) a long-term station blackout. Several sensitivity studies investigating the effect of variations in containment failure size and location, as well as hydrogen ignition concentration are also documented.

  1. Code cases for implementing risk-based inservice testing in the ASME OM code

    SciTech Connect (OSTI)

    Rowley, C.W.

    1996-12-01

    Historically inservice testing has been reasonably effective, but quite costly. Recent applications of plant PRAs to the scope of the IST program have demonstrated that of the 30 pumps and 500 valves in the typical plant IST program, less than half of the pumps and ten percent of the valves are risk significant. The way the ASME plans to tackle this overly-conservative scope for IST components is to use the PRA and plant expert panels to create a two tier IST component categorization scheme. The PRA provides the quantitative risk information and the plant expert panel blends the quantitative and deterministic information to place the IST component into one of two categories: More Safety Significant Component (MSSC) or Less Safety Significant Component (LSSC). With all the pumps and valves in the IST program placed in MSSC or LSSC categories, two different testing strategies will be applied. The testing strategies will be unique for the type of component, such as centrifugal pump, positive displacement pump, MOV, AOV, SOV, SRV, PORV, HOV, CV, and MV. A series of OM Code Cases are being developed to capture this process for a plant to use. One Code Case will be for Component Importance Ranking. The remaining Code Cases will develop the MSSC and LSSC testing strategy for type of component. These Code Cases are planned for publication in early 1997. Later, after some industry application of the Code Cases, the alternative Code Case requirements will gravitate to the ASME OM Code as appendices.

  2. MC&A System Effectiveness Tool (MSET) (Presentation 2)

    SciTech Connect (OSTI)

    Powell, Danny H; Elwood Jr, Robert H

    2011-01-01

    MSET is a self-assessment or inspection tool utilizing probabilistic risk assessment (PRA) methodology to calculate the system effectiveness of a nuclear facility's material control and accountability (MC&A) system. MSET analyzes the effectiveness of an MC&A system based on defined performance metrics for MC&A functions defined based on U.S. and international best practices and regulations. MSET analysis is based on performance of the entire MC&A system including defense-in-depth attributes and sensitivity analysis of changes in the system, both positive and negative. MSET analysis considers: accounting; containment; access control; surveillance capabilities of the system; and other interfaces with the physical protection systems that provide detection of an unauthorized action. MSET performs a system effectiveness calculation evaluation against a defined performance metric. MSET uses PRA techniques to analyze the MC&A system. MSET is a tool for evaluating the system effectiveness of MC&A systems during self-assessment or external inspection. MSET has been developed, tested, and benchmarked by the U.S. DOE. In collaboration with the U.S. DOE, Rosatom is developing a Russian version (MSET-R) planned for pilot implementation at select material balance areas in 2011. MSET has been shown to be an effective training and communication tool for MC&A.

  3. Preliminary Hazards Analysis Plasma Hearth Process

    SciTech Connect (OSTI)

    Aycock, M.; Coordes, D.; Russell, J.; TenBrook, W.; Yimbo, P.

    1993-11-01

    This Preliminary Hazards Analysis (PHA) for the Plasma Hearth Process (PHP) follows the requirements of United States Department of Energy (DOE) Order 5480.23 (DOE, 1992a), DOE Order 5480.21 (DOE, 1991d), DOE Order 5480.22 (DOE, 1992c), DOE Order 5481.1B (DOE, 1986), and the guidance provided in DOE Standards DOE-STD-1027-92 (DOE, 1992b). Consideration is given to ft proposed regulations published as 10 CFR 830 (DOE, 1993) and DOE Safety Guide SG 830.110 (DOE, 1992b). The purpose of performing a PRA is to establish an initial hazard categorization for a DOE nuclear facility and to identify those processes and structures which may have an impact on or be important to safety. The PHA is typically performed during and provides input to project conceptual design. The PRA then is followed by a Preliminary Safety Analysis Report (PSAR) performed during Title I and II design. This PSAR then leads to performance of the Final Safety Analysis Report performed during construction, testing, and acceptance and completed before routine operation. Radiological assessments indicate that a PHP facility, depending on the radioactive material inventory, may be an exempt, Category 3, or Category 2 facility. The calculated impacts would result in no significant impact to offsite personnel or the environment. Hazardous material assessments indicate that a PHP facility will be a Low Hazard facility having no significant impacts either onsite or offsite to personnel and the environment.

  4. Summary of the SRS Severe Accident Analysis Program, 1987--1992

    SciTech Connect (OSTI)

    Long, T.A.; Hyder, M.L.; Britt, T.E.; Allison, D.K.; Chow, S.; Graves, R.D.; DeWald, A.B. Jr.; Monson, P.R. Jr.; Wooten, L.A.

    1992-11-01

    The Severe Accident Analysis Program (SAAP) is a program of experimental and analytical studies aimed at characterizing severe accidents that might occur in the Savannah River Site Production Reactors. The goals of the Severe Accident Analysis Program are: To develop an understanding of severe accidents in SRS reactors that is adequate to support safety documentation for these reactors, including the Safety Analysis Report (SAR), the Probabilistic Risk Assessment (PRA), and other studies evaluating the safety of reactor operation; To provide tools and bases for the evaluation of existing or proposed safety related equipment in the SRS reactors; To provide bases for the development of accident management procedures for the SRS reactors; To develop and maintain on the site a sufficient body of knowledge, including documents, computer codes, and cognizant engineers and scientists, that can be used to authoritatively resolve questions or issues related to reactor accidents. The Severe Accident Analysis Program was instituted in 1987 and has already produced a substantial amount of information, and specialized calculational tools. Products of the Severe Accident Analysis Program (listed in Section 9 of this report) have been used in the development of the Safety Analysis Report (SAR) and the Probabilistic Risk Assessment (PRA), and in the development of technical specifications for the SRS reactors. A staff of about seven people is currently involved directly in the program and in providing input on severe accidents to other SRS activities.

  5. Code System to Calculate Integrated Reliability and Risk Analysis.

    Energy Science and Technology Software Center (OSTI)

    2002-02-18

    Version 04 IRRAS Version 4.16, the latest in a series (2.0, 2.5, 4.0, 4.15), 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 performmore » 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. IRRAS Version 4.16 is the latest in the stand-alone IRRAS series (2.0, 2.5, 4.0, 4.15). Be sure to review the PSR-405/ SAPHIRE 7.06 package which was released in January 2000 and includes three programs: the Integrated Reliability and Risk Analysis System (IRRAS), the System Analysis and Risk Assessment (SARA) system, the Models And Results Database (MAR-D) system, and the Fault tree, Event tree and P&ID (FEP) editors.« less

  6. Advance Liquid Metal Reactor Discrete Dynamic Event Tree/Bayesian Network Analysis and Incident Management Guidelines (Risk Management for Sodium Fast Reactors)

    SciTech Connect (OSTI)

    Denman, Matthew R.; Groth, Katrina M.; Cardoni, Jeffrey N.; Wheeler, Timothy A.

    2015-04-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self-correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the system's design to manage the accident. Inherently and passively safe designs are laudable, but nonetheless extreme boundary conditions can interfere with the design attributes which facilitate inherent safety, thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayesian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The authors would like to acknowledge the U.S. Department of Energy's Office of Nuclear Energy for funding this research through Work Package SR-14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at Argonne National Laboratory, Oak Ridge National Laboratory, and Idaho National Laboratory for their continue d contributions to the advanced reactor PRA mission area.

  7. Risk Management for Sodium Fast Reactors.

    SciTech Connect (OSTI)

    Denman, Matthew R; Groth, Katrina; Cardoni, Jeffrey N; Wheeler, Timothy A.

    2015-01-01

    Accident management is an important component to maintaining risk at acceptable levels for all complex systems, such as nuclear power plants. With the introduction of self - correcting, or inherently safe, reactor designs the focus has shifted from management by operators to allowing the syste m's design to manage the accident. While inherently and passively safe designs are laudable, extreme boundary conditions can interfere with the design attributes which facilitate inherent safety , thus resulting in unanticipated and undesirable end states. This report examines an inherently safe and small sodium fast reactor experiencing a beyond design basis seismic event with the intend of exploring two issues : (1) can human intervention either improve or worsen the potential end states and (2) can a Bayes ian Network be constructed to infer the state of the reactor to inform (1). ACKNOWLEDGEMENTS The author s would like to acknowledge the U.S. Department of E nergy's Office of Nuclear Energy for funding this research through Work Package SR - 14SN100303 under the Advanced Reactor Concepts program. The authors also acknowledge the PRA teams at A rgonne N ational L aborator y , O ak R idge N ational L aborator y , and I daho N ational L aborator y for their continue d contributions to the advanced reactor PRA mission area.

  8. Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) GEM Manual

    SciTech Connect (OSTI)

    C. L. Smith; J. Schroeder; S. T. Beck

    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 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). The INL's primary role in this project is that of software developer and tester. Using the SAPHIRE analysis engine and relational database is a complementary program called GEM. GEM has been designed to simplify using existing PRA analysis for activities such as the NRC’s Accident Sequence Precursor program. In this report, the theoretical framework behind GEM-type calculations are discussed in addition to providing guidance and examples for performing evaluations when using the GEM software. As part of this analysis framework, the two types of GEM analysis are outlined, specifically initiating event (where an initiator occurs) and condition (where a component is failed for some length of time) assessments.

  9. Risk-based maintenance modeling. Prioritization of maintenance importances and quantification of maintenance effectiveness

    SciTech Connect (OSTI)

    Vesely, W.E.; Rezos, J.T.

    1995-09-01

    This report describes methods for prioritizing the risk importances of maintenances using a Probabilistic Risk Assessment (PRA). Approaches then are described for quantifying their reliability and risk effects. Two different PRA importance measures, minimal cutset importances and risk reduction importances, were used to prioritize maintenances; the findings show that both give similar results if appropriate criteria are used. The justifications for the particular importance measures also are developed. The methods developed to quantify the reliability and risk effects of maintenance actions are extensions of the usual reliability models now used in PRAs. These extended models consider degraded states of the component, and quantify the benefits of maintenance in correcting degradations and preventing failures. The negative effects of maintenance, including downtimes, also are included. These models are specific types of Markov models. The data for these models can be obtained from plant maintenance logs and from the Nuclear Plant Reliability Data System (NPRDS). To explore the potential usefulness of these models, the authors analyzed a range of postulated values of input data. These models were used to examine maintenance effects on a components reliability and performance for various maintenance programs and component data. Maintenance schedules were analyzed to optimize the component`s availability. In specific cases, the effects of maintenance were found to be large.

  10. Feasibility of developing risk-based rankings of pressure boundary systems for inservice inspection

    SciTech Connect (OSTI)

    Vo, T.V.; Smith, B.W.; Simonen, F.A.; Gore, B.F.

    1994-08-01

    The goals of the Evaluation and Improvement of Non-destructive Examination Reliability for the In-service Inspection of Light Water Reactors Program sponsored by the Nuclear Regulatory Commission at Pacific Northwest Laboratory (PNL) are to (1) assess current ISI techniques and requirements for all pressure boundary systems and components, (2) determine if improvements to the requirements are needed, and (3) if necessary, develop recommendations for revising the applicable ASME Codes and regulatory requirements. In evaluating approaches that could be used to provide a technical basis for improved inservice inspection plans, PNL has developed and applied a method that uses results of probabilistic risk assessment (PRA) to establish piping system ISI requirements. In the PNL program, the feasibility of generic ISI requirements is being addressed in two phases. Phase I involves identifying and prioritizing the systems most relevant to plant safety. The results of these evaluations will be later consolidated into requirements for comprehensive inservice inspection of nuclear power plant components that will be developed in Phase II. This report presents Phase I evaluations for eight selected plants and attempts to compare these PRA-based inspection priorities with current ASME Section XI requirements for Class 1, 2 and 3 systems. These results show that there are generic insights that can be extrapolated from the selected plants to specific classes of light water reactors.

  11. SAPHIRE 8 Software Project Plan

    SciTech Connect (OSTI)

    Curtis L.Smith; Ted S. Wood

    2010-03-01

    This project is being conducted at the request of the DOE and the NRC. The INL has been requested by the NRC to improve and maintain the Systems Analysis Programs for Hands-on Integrated Reliability Evaluation (SAPHIRE) tool set concurrent with the changing needs of the user community as well as staying current with new technologies. Successful completion will be upon NRC approved release of all software and accompanying documentation in a timely fashion. This project will enhance the SAPHIRE tool set for the user community (NRC, Nuclear Power Plant operations, Probabilistic Risk Analysis (PRA) model developers) by providing improved Common Cause Failure (CCF), External Events, Level 2, and Significance Determination Process (SDP) analysis capabilities. The SAPHIRE development team at the Idaho National Laboratory is responsible for successful completion of this project. The project is under the supervision of Curtis L. Smith, PhD, Technical Lead for the SAPHIRE application. All current capabilities from SAPHIRE version 7 will be maintained in SAPHIRE 8. The following additional capabilities will be incorporated: Incorporation of SPAR models for the SDP interface. Improved quality assurance activities for PRA calculations of SAPHIRE Version 8. Continue the current activities for code maintenance, documentation, and user support for the code.

  12. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Chanin, D.I. ); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian )

    1990-02-01

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previous CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. This document, Volume 1, the Users's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems.

  13. Technical Needs for Enhancing Risk Monitors with Equipment Condition Assessment for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Ramuhalli, Pradeep; Meyer, Ryan M.; Berglin, Eric J.; Wootan, David W.; Mitchell, Mark R.

    2013-04-04

    Advanced small modular reactors (aSMRs) can provide the United States with a safe, sustainable, and carbon-neutral energy source. The controllable day-to-day costs of aSMRs are expected to be dominated by operation and maintenance costs. Health and condition assessment coupled with online risk monitors can potentially enhance affordability of aSMRs through optimized operational planning and maintenance scheduling. Currently deployed risk monitors are an extension of probabilistic risk assessment (PRA). For complex engineered systems like nuclear power plants, PRA systematically combines event likelihoods and the probability of failure (POF) of key components, so that when combined with the magnitude of possible adverse consequences to determine risk. Traditional PRA uses population-based POF information to estimate the average plant risk over time. Currently, most nuclear power plants have a PRA that reflects the as-operated, as-modified plant; this model is updated periodically, typically once a year. Risk monitors expand on living PRA by incorporating changes in the day-by-day plant operation and configuration (e.g., changes in equipment availability, operating regime, environmental conditions). However, population-based POF (or population- and time-based POF) is still used to populate fault trees. Health monitoring techniques can be used to establish condition indicators and monitoring capabilities that indicate the component-specific POF at a desired point in time (or over a desired period), which can then be incorporated in the risk monitor to provide a more accurate estimate of the plant risk in different configurations. This is particularly important for active systems, structures, and components (SSCs) proposed for use in aSMR designs. These SSCs may differ significantly from those used in the operating fleet of light-water reactors (or even in LWR-based SMR designs). Additionally, the operating characteristics of aSMRs can present significantly different

  14. EPRI/NRC-RES fire human reliability analysis guidelines.

    SciTech Connect (OSTI)

    Lewis, Stuart R.; Cooper, Susan E.; Najafi, Bijan; Collins, Erin; Hannaman, Bill; Kohlhepp, Kaydee; Grobbelaar, Jan; Hill, Kendra; Hendrickson, Stacey M. Langfitt; Forester, John Alan; Julius, Jeff

    2010-03-01

    During the 1990s, the Electric Power Research Institute (EPRI) developed methods for fire risk analysis to support its utility members in the preparation of responses to Generic Letter 88-20, Supplement 4, 'Individual Plant Examination - External Events' (IPEEE). This effort produced a Fire Risk Assessment methodology for operations at power that was used by the majority of U.S. nuclear power plants (NPPs) in support of the IPEEE program and several NPPs overseas. Although these methods were acceptable for accomplishing the objectives of the IPEEE, EPRI and the U.S. Nuclear Regulatory Commission (NRC) recognized that they required upgrades to support current requirements for risk-informed, performance-based (RI/PB) applications. In 2001, EPRI and the USNRC's Office of Nuclear Regulatory Research (RES) embarked on a cooperative project to improve the state-of-the-art in fire risk assessment to support a new risk-informed environment in fire protection. This project produced a consensus document, NUREG/CR-6850 (EPRI 1011989), entitled 'Fire PRA Methodology for Nuclear Power Facilities' which addressed fire risk for at power operations. NUREG/CR-6850 developed high level guidance on the process for identification and inclusion of human failure events (HFEs) into the fire PRA (FPRA), and a methodology for assigning quantitative screening values to these HFEs. It outlined the initial considerations of performance shaping factors (PSFs) and related fire effects that may need to be addressed in developing best-estimate human error probabilities (HEPs). However, NUREG/CR-6850 did not describe a methodology to develop best-estimate HEPs given the PSFs and the fire-related effects. In 2007, EPRI and RES embarked on another cooperative project to develop explicit guidance for estimating HEPs for human failure events under fire generated conditions, building upon existing human reliability analysis (HRA) methods. This document provides a methodology and guidance for conducting

  15. Incorporating Equipment Condition Assessment in Risk Monitors for Advanced Small Modular Reactors

    SciTech Connect (OSTI)

    Coble, Jamie B.; Coles, Garill A.; Meyer, Ryan M.; Ramuhalli, Pradeep

    2013-10-01

    Advanced small modular reactors (aSMRs) can complement the current fleet of large light-water reactors in the USA for baseload and peak demand power production and process heat applications (e.g., water desalination, shale oil extraction, hydrogen production). The day-to-day costs of aSMRs are expected to be dominated by operations and maintenance (O&M); however, the effect of diverse operating missions and unit modularity on O&M is not fully understood. These costs could potentially be reduced by optimized scheduling, with risk-informed scheduling of maintenance, repair, and replacement of equipment. Currently, most nuclear power plants have a “living” probabilistic risk assessment (PRA), which reflects the as-operated, as-modified plant and combine event probabilities with population-based probability of failure (POF) for key components. “Risk monitors” extend the PRA by incorporating the actual and dynamic plant configuration (equipment availability, operating regime, environmental conditions, etc.) into risk assessment. In fact, PRAs are more integrated into plant management in today’s nuclear power plants than at any other time in the history of nuclear power. However, population-based POF curves are still used to populate fault trees; this approach neglects the time-varying condition of equipment that is relied on during standard and non-standard configurations. Equipment condition monitoring techniques can be used to estimate the component POF. Incorporating this unit-specific estimate of POF in the risk monitor can provide a more accurate estimate of risk in different operating and maintenance configurations. This enhanced risk assessment will be especially important for aSMRs that have advanced component designs, which don’t have an available operating history to draw from, and often use passive design features, which present challenges to PRA. This paper presents the requirements and technical gaps for developing a framework to integrate unit

  16. Nuclear Material Control and Accountability System Effectiveness Tool (MSET)

    SciTech Connect (OSTI)

    Powell, Danny H; Elwood Jr, Robert H; Roche, Charles T; Campbell, Billy J; Hammond, Glenn A; Meppen, Bruce W; Brown, Richard F

    2011-01-01

    A nuclear material control and accountability (MC&A) system effectiveness tool (MSET) has been developed in the United States for use in evaluating material protection, control, and accountability (MPC&A) systems in nuclear facilities. The project was commissioned by the National Nuclear Security Administration's Office of International Material Protection and Cooperation. MSET was developed by personnel with experience spanning more than six decades in both the U.S. and international nuclear programs and with experience in probabilistic risk assessment (PRA) in the nuclear power industry. MSET offers significant potential benefits for improving nuclear safeguards and security in any nation with a nuclear program. MSET provides a design basis for developing an MC&A system at a nuclear facility that functions to protect against insider theft or diversion of nuclear materials. MSET analyzes the system and identifies several risk importance factors that show where sustainability is essential for optimal performance and where performance degradation has the greatest impact on total system risk. MSET contains five major components: (1) A functional model that shows how to design, build, implement, and operate a robust nuclear MC&A system (2) A fault tree of the operating MC&A system that adapts PRA methodology to analyze system effectiveness and give a relative risk of failure assessment of the system (3) A questionnaire used to document the facility's current MPC&A system (provides data to evaluate the quality of the system and the level of performance of each basic task performed throughout the material balance area [MBA]) (4) A formal process of applying expert judgment to convert the facility questionnaire data into numeric values representing the performance level of each basic event for use in the fault tree risk assessment calculations (5) PRA software that performs the fault tree risk assessment calculations and produces risk importance factor reports on the

  17. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    SciTech Connect (OSTI)

    Holcomb, David Eugene

    2015-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus, enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

  18. Impact of Passive Safety on FHR Instrumentation Systems Design and Classification

    SciTech Connect (OSTI)

    Holcomb, David Eugene

    2015-01-01

    Fluoride salt-cooled high-temperature reactors (FHRs) will rely more extensively on passive safety than earlier reactor classes. 10CFR50 Appendix A, General Design Criteria for Nuclear Power Plants, establishes minimum design requirements to provide reasonable assurance of adequate safety. 10CFR50.69, Risk-Informed Categorization and Treatment of Structures, Systems and Components for Nuclear Power Reactors, provides guidance on how the safety significance of systems, structures, and components (SSCs) should be reflected in their regulatory treatment. The Nuclear Energy Institute (NEI) has provided 10 CFR 50.69 SSC Categorization Guideline (NEI-00-04) that factors in probabilistic risk assessment (PRA) model insights, as well as deterministic insights, through an integrated decision-making panel. Employing the PRA to inform deterministic requirements enables an appropriately balanced, technically sound categorization to be established. No FHR currently has an adequate PRA or set of design basis accidents to enable establishing the safety classification of its SSCs. While all SSCs used to comply with the general design criteria (GDCs) will be safety related, the intent is to limit the instrumentation risk significance through effective design and reliance on inherent passive safety characteristics. For example, FHRs have no safety-significant temperature threshold phenomena, thus enabling the primary and reserve reactivity control systems required by GDC 26 to be passively, thermally triggered at temperatures well below those for which core or primary coolant boundary damage would occur. Moreover, the passive thermal triggering of the primary and reserve shutdown systems may relegate the control rod drive motors to the control system, substantially decreasing the amount of safety-significant wiring needed. Similarly, FHR decay heat removal systems are intended to be running continuously to minimize the amount of safety-significant instrumentation needed to initiate

  19. Effect of recirculation pump trip following anticipated transients without scram at Big Rock Point

    SciTech Connect (OSTI)

    Lyon, R.E.

    1981-08-01

    As requested by the US Atomic Energy Commission (now US Nuclear Regulatory Commission) in their Technical Report on Anticipated Transients Without Scram (ATWS) for Water-Cooled Reactors (WASH-1270), Consumers Power Company has submitted analyses which describe the response of their Big Rock Point (BRP) Plant to ATWS. The original analyses were submitted on Febuary 21, 1975, and results indicated that a recirculation pump trip (RPT) was effective in limiting the consequences of an ATWS. The response of BRP to an ATWS was reanalyzed as a part of the Big Rock Point Probabilistic Risk Assessment (PRA). Results of the analysis were submitted on February 26, 1981, with the conclusion that automatic RPT provides little safety improvement at BRP. Purpose of this report is to evaluate the submitted analyses to determine the effectiveness of Recirculation Pump Trip in ATWS recovery.

  20. Handbook of human-reliability analysis with emphasis on nuclear power plant applications. Final report

    SciTech Connect (OSTI)

    Swain, A D; Guttmann, H E

    1983-08-01

    The primary purpose of the Handbook is to present methods, models, and estimated human error probabilities (HEPs) to enable qualified analysts to make quantitative or qualitative assessments of occurrences of human errors in nuclear power plants (NPPs) that affect the availability or operational reliability of engineered safety features and components. The Handbook is intended to provide much of the modeling and information necessary for the performance of human reliability analysis (HRA) as a part of probabilistic risk assessment (PRA) of NPPs. Although not a design guide, a second purpose of the Handbook is to enable the user to recognize error-likely equipment design, plant policies and practices, written procedures, and other human factors problems so that improvements can be considered. The Handbook provides the methodology to identify and quantify the potential for human error in NPP tasks.

  1. RAVEN as a tool for dynamic probabilistic risk assessment: Software overview

    SciTech Connect (OSTI)

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

    2013-07-01

    RAVEN is a software tool under development at the Idaho National Laboratory (INL) that acts as the control logic driver and post-processing tool for the newly developed Thermal-Hydraulic code RELAP-7. The scope of this paper is to show the software structure of RAVEN and its utilization in connection with RELAP-7. A short overview of the mathematical framework behind the code is presented along with its main capabilities such as on-line controlling/ monitoring and Monte-Carlo sampling. A demo of a Station Black Out PRA analysis of a simplified Pressurized Water Reactor (PWR) model is shown in order to demonstrate the Monte-Carlo and clustering capabilities. (authors)

  2. Issues in benchmarking human reliability analysis methods : a literature review.

    SciTech Connect (OSTI)

    Lois, Erasmia; Forester, John Alan; Tran, Tuan Q.; Hendrickson, Stacey M. Langfitt; Boring, Ronald L.

    2008-04-01

    There is a diversity of human reliability analysis (HRA) methods available for use in assessing human performance within probabilistic risk assessment (PRA). Due to the significant differences in the methods, including the scope, approach, and underlying models, there is a need for an empirical comparison investigating the validity and reliability of the methods. To accomplish this empirical comparison, a benchmarking study is currently underway that compares HRA methods with each other and against operator performance in simulator studies. In order to account for as many effects as possible in the construction of this benchmarking study, a literature review was conducted, reviewing past benchmarking studies in the areas of psychology and risk assessment. A number of lessons learned through these studies are presented in order to aid in the design of future HRA benchmarking endeavors.

  3. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J.

    1993-11-01

    Results are presented of a systems failure analysis of the PIUS plant systems that are used during normal reactor operation and postulated accidents. This study was performed to provide the NRC with an understanding of the behavior of the plant. The study applied two diverse failure identification methods, Failure Modes Effects & Criticality Analysis (FMECA) and Hazards & Operability (HAZOP) to the plant systems, supported by several deterministic analyses. Conventional PRA methods were also used along with a scheme for classifying events by initiator frequency and combinations of failures. Principal results of this study are: (a) an extensive listing of potential event sequences, grouped in categories that can be used by the NRC, (b) identification of support systems that are important to safety, and (c) identification of key operator actions.

  4. Grid Inertial Response-Based Probabilistic Determination of Energy Storage System Capacity Under High Solar Penetration

    SciTech Connect (OSTI)

    Yue, Meng; Wang, Xiaoyu

    2015-07-01

    It is well-known that responsive battery energy storage systems (BESSs) are an effective means to improve the grid inertial response to various disturbances including the variability of the renewable generation. One of the major issues associated with its implementation is the difficulty in determining the required BESS capacity mainly due to the large amount of inherent uncertainties that cannot be accounted for deterministically. In this study, a probabilistic approach is proposed to properly size the BESS from the perspective of the system inertial response, as an application of probabilistic risk assessment (PRA). The proposed approach enables a risk-informed decision-making process regarding (1) the acceptable level of solar penetration in a given system and (2) the desired BESS capacity (and minimum cost) to achieve an acceptable grid inertial response with a certain confidence level.

  5. Performing Probabilistic Risk Assessment Through RAVEN

    SciTech Connect (OSTI)

    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

  6. Grid Inertial Response-Based Probabilistic Determination of Energy Storage System Capacity Under High Solar Penetration

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Yue, Meng; Wang, Xiaoyu

    2015-07-01

    It is well-known that responsive battery energy storage systems (BESSs) are an effective means to improve the grid inertial response to various disturbances including the variability of the renewable generation. One of the major issues associated with its implementation is the difficulty in determining the required BESS capacity mainly due to the large amount of inherent uncertainties that cannot be accounted for deterministically. In this study, a probabilistic approach is proposed to properly size the BESS from the perspective of the system inertial response, as an application of probabilistic risk assessment (PRA). The proposed approach enables a risk-informed decision-making processmore » regarding (1) the acceptable level of solar penetration in a given system and (2) the desired BESS capacity (and minimum cost) to achieve an acceptable grid inertial response with a certain confidence level.« less

  7. Exploratory Nuclear Reactor Safety Analysis and Visualization via Integrated Topological and Geometric Techniques

    SciTech Connect (OSTI)

    Dan Maljovec; Bei Wang; Valerio Pascucci; Peer-Timo Bremer; Diego Mandelli; Michael Pernice; Robert Nourgaliev

    2013-10-01

    A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis

  8. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events

    SciTech Connect (OSTI)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L.; Sandia National Labs., Albuquerque, NM )

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. 58 refs., 58 figs., 52 tabs.

  9. New time-line technique for station blackout core-melt analysis

    SciTech Connect (OSTI)

    Stutzke, M.A.

    1986-01-01

    Florida Power Corporation (FPC) has developed a new method for analyzing station blackout (SBO) core-melt accidents. This method, created during the recent probabilistic risk assessment (PRA) of Crystal River Unit 3 (CR-3), originated from the need to analyze the interactions among the two-train emergency feedwater (EFW) system, station batteries, and diesel generators (DGs) following a loss of off-site power (LOSP) event. SBO core-melt sequences for CR-3 are unique since the time core-melt commences depends on which DG fails last. The purpose of this paper is to outline the new method of analysis of SBO core-melt accidents at CR-3. The significance of SBO core-melt accidents to total plant risk, along with the efficacy of various methods to reduce SBO risk, are also discussed.

  10. Risk-based inspection guide for Crystal River Unit 3 Nuclear Power Plant

    SciTech Connect (OSTI)

    Smith, B.W.; Dukelow, J.S.; Vo, T.V.; Harris, M.S.; Gore, B.F.; Hunt, S.T. )

    1991-06-01

    The Level 1 probabilistic risk assessment (PRA) for Crystal River Unit 3 (CR-3) has been analyzed to identify plant systems and components important to minimizing public risk, as measured by system contributions to plant core damage frequency, and to identify the primary failure modes for these components. The report presents a series of tables, organized by system and prioritized by risk importance, which identify components associated with 98% of the inspectable risk due to plant operation. The systems addressed, in descending order to risk importance are: Low Pressure Injection, AC Power, Service Water, Demineralized Water, High Pressure Injection, DC Power, Emergency Feedwater, Reactor Coolant Pressure Control, and Power Conversion. This ranking is based on the Fussell-Vesely measure of risk importance, i.e., the fraction of the total core damage frequency which involves failures of the system of interest. 3 refs., 9 figs., 13 tabs.

  11. Human Events Reference for ATHEANA (HERA) Database Description and Preliminary User's Manual

    SciTech Connect (OSTI)

    Auflick, J.L.

    1999-08-12

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database (db) of analytical operational events, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA.

  12. Human events reference for ATHEANA (HERA) database description and preliminary user`s manual

    SciTech Connect (OSTI)

    Auflick, J.L.; Hahn, H.A.; Pond, D.J.

    1998-05-27

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavioral science models. The fundamental premise of ATHEANA is that error-forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. This report documents the initial development efforts for HERA.

  13. Modeling the thermal and structural response of engineered systems to abnormal environments

    SciTech Connect (OSTI)

    Skocypec, R.D.; Thomas, R.K.; Moya, J.L.

    1993-10-01

    Sandia National Laboratories (SNL) is engaged actively in research to improve the ability to accurately predict the response of engineered systems to thermal and structural abnormal environments. Abnormal environments that will be addressed in this paper include: fire, impact, and puncture by probes and fragments, as well as a combination of all of the above. Historically, SNL has demonstrated the survivability of engineered systems to abnormal environments using a balanced approach between numerical simulation and testing. It is necessary to determine the response of engineered systems in two cases: (1) to satisfy regulatory specifications, and (2) to enable quantification of a probabilistic risk assessment (PRA). In a regulatory case, numerical simulation of system response is generally used to guide the system design such that the system will respond satisfactorily to the specified regulatory abnormal environment. Testing is conducted at the regulatory abnormal environment to ensure compliance.

  14. Risk-Informing Safety Reviews for Non-Reactor Nuclear Facilities

    SciTech Connect (OSTI)

    Mubayi, V.; Azarm, A.; Yue, M.; Mukaddam, W.; Good, G.; Gonzalez, F.; Bari, R.A.

    2011-03-13

    This paper describes a methodology used to model potential accidents in fuel cycle facilities that employ chemical processes to separate and purify nuclear materials. The methodology is illustrated with an example that uses event and fault trees to estimate the frequency of a specific energetic reaction that can occur in nuclear material processing facilities. The methodology used probabilistic risk assessment (PRA)-related tools as well as information about the chemical reaction characteristics, information on plant design and operational features, and generic data about component failure rates and human error rates. The accident frequency estimates for the specific reaction help to risk-inform the safety review process and assess compliance with regulatory requirements.

  15. RAVEN AS A TOOL FOR DYNAMIC PROBABILISTIC RISK ASSESSMENT: SOFTWARE OVERVIEW

    SciTech Connect (OSTI)

    Alfonsi Andrea; Mandelli Diego; Rabiti Cristian; Joshua Cogliati; Robert Kinoshita

    2013-05-01

    RAVEN is a software tool under development at the Idaho National Laboratory (INL) that acts as the control logic driver and post-processing tool for the newly developed Thermo-Hydraylic code RELAP- 7. The scope of this paper is to show the software structure of RAVEN and its utilization in connection with RELAP-7. A short overview of the mathematical framework behind the code is presented along with its main capabilities such as on-line controlling/monitoring and Monte-Carlo sampling. A demo of a Station Black Out PRA analysis of a simplified Pressurized Water Reactor (PWR) model is shown in order to demonstrate the Monte-Carlo and clustering capabilities.

  16. The primary test of measuremental system for the actual emittance of relativistic electron beams

    SciTech Connect (OSTI)

    Liang Fu; Tai-bin Du; Xin Chen

    1995-12-31

    Recent, a new measuremental system has been established basically in Tsinghua University PRA. This system is able to measure the lower emittance of the electron beams from the RF accelerators for the FEL. It consists of a scanning magnetic field, a slit, a fluorescent screen, and a TV camera, an image processing system, a CAD 386 computer. Using it an actual phase diagram is obtained for 4-10 Mev electron beams, The principle and structure of the facility were reported in the Proceeding of the 15th FEL Conference. This paper describes the performance of the main components and the results of first measurement for the electron gun and 4Mev standing wave LINAC, Some new suggests are related too.

  17. ROBUSTNESS OF DECISION INSIGHTS UNDER ALTERNATIVE ALEATORY/EPISTEMIC UNCERTAINTY CLASSIFICATIONS

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Eslinger, Paul W.; Johnson, Kenneth I.

    2013-09-22

    The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). A key technical challenge is to establish the conceptual and technical feasibility of analyzing safety margin in a risk-informed way, which, unlike conventionally defined deterministic margin analysis, would be founded on probabilistic characterizations of SSC performance. Evaluation of probabilistic safety margins will in general entail the uncertainty characterization both of the prospective challenge to the performance of an SSC ("load") and of its "capacity" to withstand that challenge. The RISMC framework contrasts sharply with the traditional probabilistic risk assessment (PRA) structure in that the underlying models are not inherently aleatory. Rather, they are largely deterministic physical/engineering models with ambiguities about the appropriate uncertainty classification of many model parameters. The current analysis demonstrates that if the distinction between epistemic and aleatory uncertainties is to be preserved in a RISMC-like modeling environment, then it is unlikely that analysis insights supporting decision-making will in general be robust under recategorization of input uncertainties. If it is believed there is a true conceptual distinction between epistemic and aleatory uncertainty (as opposed to the distinction being primarily a legacy of the PRA paradigm) then a consistent and defensible basis must be established by which to categorize input uncertainties.

  18. Reduced Order Model Implementation in the Risk-Informed Safety Margin Characterization Toolkit

    SciTech Connect (OSTI)

    Mandelli, Diego; Smith, Curtis L.; Alfonsi, Andrea; Rabiti, Cristian; Cogliati, Joshua J.; Talbot, Paul W.; Rinaldi, Ivan; Maljovec, Dan; Wang, Bei; Pascucci, Valerio; Zhao, Haihua

    2015-09-01

    The RISMC project aims to develop new advanced simulation-based tools to perform Probabilistic Risk Analysis (PRA) for the existing fleet of U.S. nuclear power plants (NPPs). These tools numerically model not only the thermo-hydraulic behavior of the reactor primary and secondary systems but also external events temporal evolution and components/system ageing. Thus, this is not only a multi-physics problem but also a multi-scale problem (both spatial, µm-mm-m, and temporal, ms-s-minutes-years). As part of the RISMC PRA approach, a large amount of computationally expensive simulation runs are required. An important aspect is that even though computational power is regularly growing, the overall computational cost of a RISMC analysis may be not viable for certain cases. A solution that is being evaluated is the use of reduce order modeling techniques. During the FY2015, we investigated and applied reduced order modeling techniques to decrease the RICM analysis computational cost by decreasing the number of simulations runs to perform and employ surrogate models instead of the actual simulation codes. This report focuses on the use of reduced order modeling techniques that can be applied to any RISMC analysis to generate, analyze and visualize data. In particular, we focus on surrogate models that approximate the simulation results but in a much faster time (µs instead of hours/days). We apply reduced order and surrogate modeling techniques to several RISMC types of analyses using RAVEN and RELAP-7 and show the advantages that can be gained.

  19. Analysis of the Space Propulsion System Problem Using RAVEN

    SciTech Connect (OSTI)

    diego mandelli; curtis smith; cristian rabiti; andrea alfonsi

    2014-06-01

    This paper presents the solution of the space propulsion problem using a PRA code currently under development at Idaho National Laboratory (INL). RAVEN (Reactor Analysis and Virtual control ENviroment) is a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities. It is designed to derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures) and to perform both Monte- Carlo sampling of random distributed events and Event Tree based analysis. In order to facilitate the input/output handling, a Graphical User Interface (GUI) and a post-processing data-mining module are available. RAVEN allows also to interface with several numerical codes such as RELAP5 and RELAP-7 and ad-hoc system simulators. For the space propulsion system problem, an ad-hoc simulator has been developed and written in python language and then interfaced to RAVEN. Such simulator fully models both deterministic (e.g., system dynamics and interactions between system components) and stochastic behaviors (i.e., failures of components/systems such as distribution lines and thrusters). Stochastic analysis is performed using random sampling based methodologies (i.e., Monte-Carlo). Such analysis is accomplished to determine both the reliability of the space propulsion system and to propagate the uncertainties associated to a specific set of parameters. As also indicated in the scope of the benchmark problem, the results generated by the stochastic analysis are used to generate risk-informed insights such as conditions under witch different strategy can be followed.

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

    SciTech Connect (OSTI)

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

    1994-06-01

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

  1. RAVEN. Dynamic Event Tree Approach Level III Milestone

    SciTech Connect (OSTI)

    Alfonsi, Andrea; Rabiti, Cristian; Mandelli, Diego; Cogliati, Joshua; Kinoshita, Robert

    2014-07-01

    Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics are not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (DPRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed to perform two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, the control logic infrastructure is used to model stochastic events, such as components failures, and perform uncertainty propagation. Such stochastic modeling is deployed using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This report focuses on the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, a DPRA analysis, using DET, of a simplified pressurized water reactor for a Station Black-Out (SBO) scenario is presented.

  2. Optimized Uncertainty Quantification Algorithm Within a Dynamic Event Tree Framework

    SciTech Connect (OSTI)

    J. W. Nielsen; Akira Tokuhiro; Robert Hiromoto

    2014-06-01

    Methods for developing Phenomenological Identification and Ranking Tables (PIRT) for nuclear power plants have been a useful tool in providing insight into modelling aspects that are important to safety. These methods have involved expert knowledge with regards to reactor plant transients and thermal-hydraulic codes to identify are of highest importance. Quantified PIRT provides for rigorous method for quantifying the phenomena that can have the greatest impact. The transients that are evaluated and the timing of those events are typically developed in collaboration with the Probabilistic Risk Analysis. Though quite effective in evaluating risk, traditional PRA methods lack the capability to evaluate complex dynamic systems where end states may vary as a function of transition time from physical state to physical state . Dynamic PRA (DPRA) methods provide a more rigorous analysis of complex dynamic systems. A limitation of DPRA is its potential for state or combinatorial explosion that grows as a function of the number of components; as well as, the sampling of transition times from state-to-state of the entire system. This paper presents a method for performing QPIRT within a dynamic event tree framework such that timing events which result in the highest probabilities of failure are captured and a QPIRT is performed simultaneously while performing a discrete dynamic event tree evaluation. The resulting simulation results in a formal QPIRT for each end state. The use of dynamic event trees results in state explosion as the number of possible component states increases. This paper utilizes a branch and bound algorithm to optimize the solution of the dynamic event trees. The paper summarizes the methods used to implement the branch-and-bound algorithm in solving the discrete dynamic event trees.

  3. Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models

    SciTech Connect (OSTI)

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

    2014-07-30

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

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

    SciTech Connect (OSTI)

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

    1994-06-01

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

  5. RAVEN: Dynamic Event Tree Approach Level III Milestone

    SciTech Connect (OSTI)

    Andrea Alfonsi; Cristian Rabiti; Diego Mandelli; Joshua Cogliati; Robert Kinoshita

    2013-07-01

    Conventional Event-Tree (ET) based methodologies are extensively used as tools to perform reliability and safety assessment of complex and critical engineering systems. One of the disadvantages of these methods is that timing/sequencing of events and system dynamics are not explicitly accounted for in the analysis. In order to overcome these limitations several techniques, also know as Dynamic Probabilistic Risk Assessment (DPRA), have been developed. Monte-Carlo (MC) and Dynamic Event Tree (DET) are two of the most widely used D-PRA methodologies to perform safety assessment of Nuclear Power Plants (NPP). In the past two years, the Idaho National Laboratory (INL) has developed its own tool to perform Dynamic PRA: RAVEN (Reactor Analysis and Virtual control ENvironment). RAVEN has been designed to perform two main tasks: 1) control logic driver for the new Thermo-Hydraulic code RELAP-7 and 2) post-processing tool. In the first task, RAVEN acts as a deterministic controller in which the set of control logic laws (user defined) monitors the RELAP-7 simulation and controls the activation of specific systems. Moreover, the control logic infrastructure is used to model stochastic events, such as components failures, and perform uncertainty propagation. Such stochastic modeling is deployed using both MC and DET algorithms. In the second task, RAVEN processes the large amount of data generated by RELAP-7 using data-mining based algorithms. This report focuses on the analysis of dynamic stochastic systems using the newly developed RAVEN DET capability. As an example, a DPRA analysis, using DET, of a simplified pressurized water reactor for a Station Black-Out (SBO) scenario is presented.

  6. Advanced Test Reactor outage risk assessment

    SciTech Connect (OSTI)

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

    1997-12-31

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

  7. RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA

    SciTech Connect (OSTI)

    Carelli, M.D.; Petrovic, B.

    2004-10-03

    The IRIS (International Reactor Innovative and Secure) reactor design is being developed by an international consortium of 21 organizations from ten countries, including three members from Brazil and one from Mexico. This reflects the interest that Latin America has for a project which addresses the energy needs of the region. Presented here are some of the most recent developments in the IRIS project. The project's highest priority is the current pre-application licensing with the US NRC, which has required an investigation of the major accident sequences and a preliminary probabilistic risk assessment (PRA). The results of the accident analyses confirmed the outstanding inherent safety of the IRIS configuration and the PRA analyses indicated a core damage frequency due to internal events of the order of 2E-8. This not only highlights the enhanced safety characteristic of IRIS which should enhance its public acceptance, but it has also prompted IRIS to consider the possibility of being licensed without the need for off-site emergency response planning which would have a very positive economic implication. The modular IRIS, with each module rated at {approx} 335 MWe, is of course an ideal size for developing countries as it allows to easily introduce a moderate amount of power on limited electric grids. IRIS can be deployed in single modules in regions only requiring a few hundred MWs or in multiple modules deployed successively at time intervals in large urban areas requiring a larger amount of power increasing with time. IRIS is designed to operate ''hands-off'' as much as possible, with a small crew, having in mind deployment in areas with limited infrastructure. Thus IRIS has a 48-months maintenance interval, long refueling cycles in excess of three years, and is designed to increase as much as possible operational reliability. For example, the project has recently adopted internal control rod drive mechanisms to eliminate vessel head penetrations and the

  8. Conversion of Questionnaire Data

    SciTech Connect (OSTI)

    Powell, Danny H; Elwood Jr, Robert H

    2011-01-01

    During the survey, respondents are asked to provide qualitative answers (well, adequate, needs improvement) on how well material control and accountability (MC&A) functions are being performed. These responses can be used to develop failure probabilities for basic events performed during routine operation of the MC&A systems. The failure frequencies for individual events may be used to estimate total system effectiveness using a fault tree in a probabilistic risk analysis (PRA). Numeric risk values are required for the PRA fault tree calculations that are performed to evaluate system effectiveness. So, the performance ratings in the questionnaire must be converted to relative risk values for all of the basic MC&A tasks performed in the facility. If a specific material protection, control, and accountability (MPC&A) task is being performed at the 'perfect' level, the task is considered to have a near zero risk of failure. If the task is performed at a less than perfect level, the deficiency in performance represents some risk of failure for the event. As the degree of deficiency in performance increases, the risk of failure increases. If a task that should be performed is not being performed, that task is in a state of failure. The failure probabilities of all basic events contribute to the total system risk. Conversion of questionnaire MPC&A system performance data to numeric values is a separate function from the process of completing the questionnaire. When specific questions in the questionnaire are answered, the focus is on correctly assessing and reporting, in an adjectival manner, the actual performance of the related MC&A function. Prior to conversion, consideration should not be given to the numeric value that will be assigned during the conversion process. In the conversion process, adjectival responses to questions on system performance are quantified based on a log normal scale typically used in human error analysis (see A.D. Swain and H.E. Guttmann

  9. Development and application of the dynamic system doctor to nuclear reactor probabilistic risk assessments.

    SciTech Connect (OSTI)

    Kunsman, David Marvin; Aldemir, Tunc; Rutt, Benjamin; Metzroth, Kyle; Catalyurek, Umit; Denning, Richard; Hakobyan, Aram; Dunagan, Sean C.

    2008-05-01

    This LDRD project has produced a tool that makes probabilistic risk assessments (PRAs) of nuclear reactors - analyses which are very resource intensive - more efficient. PRAs of nuclear reactors are being increasingly relied on by the United States Nuclear Regulatory Commission (U.S.N.R.C.) for licensing decisions for current and advanced reactors. Yet, PRAs are produced much as they were 20 years ago. The work here applied a modern systems analysis technique to the accident progression analysis portion of the PRA; the technique was a system-independent multi-task computer driver routine. Initially, the objective of the work was to fuse the accident progression event tree (APET) portion of a PRA to the dynamic system doctor (DSD) created by Ohio State University. Instead, during the initial efforts, it was found that the DSD could be linked directly to a detailed accident progression phenomenological simulation code - the type on which APET construction and analysis relies, albeit indirectly - and thereby directly create and analyze the APET. The expanded DSD computational architecture and infrastructure that was created during this effort is called ADAPT (Analysis of Dynamic Accident Progression Trees). ADAPT is a system software infrastructure that supports execution and analysis of multiple dynamic event-tree simulations on distributed environments. A simulator abstraction layer was developed, and a generic driver was implemented for executing simulators on a distributed environment. As a demonstration of the use of the methodological tool, ADAPT was applied to quantify the likelihood of competing accident progression pathways occurring for a particular accident scenario in a particular reactor type using MELCOR, an integrated severe accident analysis code developed at Sandia. (ADAPT was intentionally created with flexibility, however, and is not limited to interacting with only one code. With minor coding changes to input files, ADAPT can be linked to other

  10. Component Repair Times Obtained from MSPI Data

    SciTech Connect (OSTI)

    Eide, Steven A.

    2015-05-01

    Information concerning times to repair or restore equipment to service given a failure is valuable to probabilistic risk assessments (PRAs). Examples of such uses in modern PRAs include estimation of the probability of failing to restore a failed component within a specified time period (typically tied to recovering a mitigating system before core damage occurs at nuclear power plants) and the determination of mission times for support system initiating event (SSIE) fault tree models. Information on equipment repair or restoration times applicable to PRA modeling is limited and dated for U.S. commercial nuclear power plants. However, the Mitigating Systems Performance Index (MSPI) program covering all U.S. commercial nuclear power plants provides up-to-date information on restoration times for a limited set of component types. This paper describes the MSPI program data available and analyzes the data to obtain median and mean component restoration times as well as non-restoration cumulative probability curves. The MSPI program provides guidance for monitoring both planned and unplanned outages of trains of selected mitigating systems deemed important to safety. For systems included within the MSPI program, plants monitor both train UA and component unreliability (UR) against baseline values. If the combined system UA and UR increases sufficiently above established baseline results (converted to an estimated change in core damage frequency or CDF), a “white” (or worse) indicator is generated for that system. That in turn results in increased oversight by the US Nuclear Regulatory Commission (NRC) and can impact a plant’s insurance rating. Therefore, there is pressure to return MSPI program components to service as soon as possible after a failure occurs. Three sets of unplanned outages might be used to determine the component repair durations desired in this article: all unplanned outages for the train type that includes the component of interest, only

  11. Insights from an overview of four PRAs

    SciTech Connect (OSTI)

    Fitzpatrick, R.; Arrieta, L.; Teichmann, T.; Davis, P.

    1986-01-01

    This paper summarizes the findings of an investigation of four probabilistic risk assessments (PRAs), those for Millstone 3, Seabrook, Shoreham, and Oconee 3, performed by Brookhaven National Laboratory (BNL) for the Reliability and Risk Assessment Branch of the US Nuclear Regulatory Commission (NRC). This group of four PRAs was subjected to an overview process with the basic goal of ascertaining what insights might be gained (beyond those already documented within the individual PRAs) by an independent evaluation of the group with respect to nuclear plant safety and vulnerability. Specifically, the objectives of the study were (1) to identify and rank initiators, systems, components, and failure modes from dominant accident sequences according to their contribution to core melt probability and public risk; and (2) to derive from this process plant-specific and generic insights. The effort was not intended to verify the specific details and results of each PRA but rather - having accepted the results - to see what they might mean in a more global context. The NRC had previously sponsored full detailed reviews of each of these PRAs, but only two, those for Millstone 3 and Shoreham, were completed and documented in time to allow their consideration within the study. This paper also presents some comments and insights into the amenability of certain features of these PRAs to this type of overview process.

  12. RAMONA-3B application to Browns Ferry ATWS

    SciTech Connect (OSTI)

    Slovik, G.C.; Neymotin, L.Y.; Saha, P.

    1985-01-01

    The Anticipated Transient Without Scram (ATWS) is known to be a dominant accident sequence for possible core melt in a Boiling Water Reactor (BWR). A recent Probabilistic Risk Assessment (PRA) analysis for the Browns Ferry nuclear power plant indicates that ATWS is the second most dominant transient for core melt in BWR/4 with Mark I containment. The most dominant sequence being the failure of long term decay heat removal function of the Residual Heat Removal (RHR) system. Of all the various ATWS scenarios, the Main Steam Isolation Valve (MSIV) closure ATWS sequence was chosen for present analysis because of its relatively high frequency of occurrence and its challenge to the residual heat removal system and containment integrity. The objective of this paper is to discuss four MSIV closure ATWS calculations using the RAMONA-3B code. The paper is a summary of a report being prepared for the USNRC Severe Accident Sequence Analysis (SASA) program which should be referred to for details. 10 refs., 20 figs., 3 tabs.

  13. A Bayesian method for using simulator data to enhance human error probabilities assigned by existing HRA methods

    DOE Public Access Gateway for Energy & Science Beta (PAGES Beta)

    Groth, Katrina M.; Smith, Curtis L.; Swiler, Laura P.

    2014-04-05

    In the past several years, several international agencies have begun to collect data on human performance in nuclear power plant simulators [1]. This data provides a valuable opportunity to improve human reliability analysis (HRA), but there improvements will not be realized without implementation of Bayesian methods. Bayesian methods are widely used in to incorporate sparse data into models in many parts of probabilistic risk assessment (PRA), but Bayesian methods have not been adopted by the HRA community. In this article, we provide a Bayesian methodology to formally use simulator data to refine the human error probabilities (HEPs) assigned by existingmore » HRA methods. We demonstrate the methodology with a case study, wherein we use simulator data from the Halden Reactor Project to update the probability assignments from the SPAR-H method. The case study demonstrates the ability to use performance data, even sparse data, to improve existing HRA methods. Furthermore, this paper also serves as a demonstration of the value of Bayesian methods to improve the technical basis of HRA.« less

  14. Development of Simplified Probabilistic Risk Assessment Model for Seismic Initiating Event

    SciTech Connect (OSTI)

    S. Khericha; R. Buell; S. Sancaktar; M. Gonzalez; F. Ferrante

    2012-06-01

    ABSTRACT This paper discusses a simplified method to evaluate seismic risk using a methodology built on dividing the seismic intensity spectrum into multiple discrete bins. The seismic probabilistic risk assessment model uses Nuclear Regulatory Commissions (NRCs) full power Standardized Plant Analysis Risk (SPAR) model as the starting point for development. The seismic PRA models are integrated with their respective internal events at-power SPAR model. This is accomplished by combining the modified system fault trees from the full power SPAR model with seismic event tree logic. The peak ground acceleration is divided into five bins. The g-value for each bin is estimated using the geometric mean of lower and upper values of that particular bin and the associated frequency for each bin is estimated by taking the difference between upper and lower values of that bin. The components fragilities are calculated for each bin using the plant data, if available, or generic values of median peak ground acceleration and uncertainty values for the components. For human reliability analysis (HRA), the SPAR HRA (SPAR-H) method is used which requires the analysts to complete relatively straight forward worksheets that include the performance shaping factors (PSFs). The results are then used to estimate human error probabilities (HEPs) of interest. This work is expected to improve the NRCs ability to include seismic hazards in risk assessments for operational events in support of the reactor oversight program (e.g., significance determination process).

  15. Development of Probabilistic Risk Assessment Model for BWR Shutdown Modes 4 and 5 Integrated in SPAR Model

    SciTech Connect (OSTI)

    S. T. Khericha; S. Sancakter; J. Mitman; J. Wood

    2010-06-01

    Nuclear plant operating experience and several studies show that the risk from shutdown operation during modes 4, 5, and 6 can be significant This paper describes development of the standard template risk evaluation models for shutdown modes 4, and 5 for commercial boiling water nuclear power plants (BWR). The shutdown probabilistic risk assessment model uses full power Nuclear Regulatory Commissions (NRCs) Standardized Plant Analysis Risk (SPAR) model as the starting point for development. The shutdown PRA models are integrated with their respective internal events at-power SPAR model. This is accomplished by combining the modified system fault trees from SPAR full power model with shutdown event tree logic. For human reliability analysis (HRA), the SPAR HRA (SPAR-H) method is used which requires the analysts to complete relatively straight forward worksheet, including the performance shaping factors (PSFs). The results are then used to estimate HEP of interest. The preliminary results indicate the risk is dominated by the operators ability to diagnose the events and provide long term cooling.

  16. Risk assessment handbook

    SciTech Connect (OSTI)

    Farmer, F.G.; Jones, J.L.; Hunt, R.N.; Roush, M.L.; Wierman, T.E.

    1990-09-01

    The Probabilistic Risk Assessment Unit at EG G Idaho has developed this handbook to provide guidance to a facility manager exploring the potential benefit to be gained by performance of a risk assessment properly scoped to meet local needs. This document is designed to help the manager control the resources expended commensurate with the risks being managed and to assure that the products can be used programmatically to support future needs in order to derive maximum beneflt from the resources expended. We present a logical and functional mapping scheme between several discrete phases of project definition to ensure that a potential customer, working with an analyst, is able to define the areas of interest and that appropriate methods are employed in the analysis. In addition the handbook is written to provide a high-level perspective for the analyst. Previously, the needed information was either scattered or existed only in the minds of experienced analysts. By compiling this information and exploring the breadth of knowledge which exists within the members of the PRA Unit, the functional relationships between the customers' needs and the product have been established.

  17. Risk assessment handbook

    SciTech Connect (OSTI)

    Farmer, F.G.; Jones, J.L.; Hunt, R.N.; Roush, M.L.; Wierman, T.E.

    1990-09-01

    The Probabilistic Risk Assessment Unit at EG&G Idaho has developed this handbook to provide guidance to a facility manager exploring the potential benefit to be gained by performance of a risk assessment properly scoped to meet local needs. This document is designed to help the manager control the resources expended commensurate with the risks being managed and to assure that the products can be used programmatically to support future needs in order to derive maximum beneflt from the resources expended. We present a logical and functional mapping scheme between several discrete phases of project definition to ensure that a potential customer, working with an analyst, is able to define the areas of interest and that appropriate methods are employed in the analysis. In addition the handbook is written to provide a high-level perspective for the analyst. Previously, the needed information was either scattered or existed only in the minds of experienced analysts. By compiling this information and exploring the breadth of knowledge which exists within the members of the PRA Unit, the functional relationships between the customers` needs and the product have been established.

  18. FATE Unified Modeling Method for Spent Nuclear Fuel and Sludge Processing, Shipping and Storage - 13405

    SciTech Connect (OSTI)

    Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert

    2013-07-01

    A unified modeling method applicable to the processing, shipping, and storage of spent nuclear fuel and sludge has been incrementally developed, validated, and applied over a period of about 15 years at the US DOE Hanford site. The software, FATE{sup TM}, provides a consistent framework for a wide dynamic range of common DOE and commercial fuel and waste applications. It has been used during the design phase, for safety and licensing calculations, and offers a graded approach to complex modeling problems encountered at DOE facilities and abroad (e.g., Sellafield). FATE has also been used for commercial power plant evaluations including reactor building fire modeling for fire PRA, evaluation of hydrogen release, transport, and flammability for post-Fukushima vulnerability assessment, and drying of commercial oxide fuel. FATE comprises an integrated set of models for fluid flow, aerosol and contamination release, transport, and deposition, thermal response including chemical reactions, and evaluation of fire and explosion hazards. It is one of few software tools that combine both source term and thermal-hydraulic capability. Practical examples are described below, with consideration of appropriate model complexity and validation. (authors)

  19. Research prioritization using the Analytic Hierarchy Process: basic methods. Volume 1

    SciTech Connect (OSTI)

    Vesely, W.E.; Shafaghi, A.; Gary, I. Jr.; Rasmuson, D.M.

    1983-08-01

    This report describes a systematic approach for prioritizing research needs and research programs. The approach is formally called the Analytic Hierarchy Process which was developed by T.L. Saaty and is described in several of his texts referenced in the report. The Analytic Hierarchy Process, or AHP for short, has been applied to a wide variety of prioritization problems and has a good record of success as documented in Saaty's texts. The report develops specific guidelines for constructing the hierarchy and for prioritizing the research programs. Specific examples are given to illustrate the steps in the AHP. As part of the work, a computer code has been developed and the use of the code is described. The code allows the prioritizations to be done in a codified and efficient manner; sensitivity and parametric studies can also be straightforwardly performed to gain a better understanding of the prioritization results. Finally, as an important part of the work, an approach is developed which utilizes probabilistic risk analyses (PRAs) to systematically identify and prioritize research needs and research programs. When utilized in an AHP framework, the PRA's which have been performed to date provide a powerful information source for focusing research on those areas most impacting risk and risk uncertainty.

  20. A Bayesian method for using simulator data to enhance human error probabilities assigned by existing HRA methods

    SciTech Connect (OSTI)

    Katrinia M. Groth; Curtis L. Smith; Laura P. Swiler

    2014-08-01

    In the past several years, several international organizations have begun to collect data on human performance in nuclear power plant simulators. The data collected provide a valuable opportunity to improve human reliability analysis (HRA), but these improvements will not be realized without implementation of Bayesian methods. Bayesian methods are widely used to incorporate sparse data into models in many parts of probabilistic risk assessment (PRA), but Bayesian methods have not been adopted by the HRA community. In this paper, we provide a Bayesian methodology to formally use simulator data to refine the human error probabilities (HEPs) assigned by existing HRA methods. We demonstrate the methodology with a case study, wherein we use simulator data from the Halden Reactor Project to update the probability assignments from the SPAR-H method. The case study demonstrates the ability to use performance data, even sparse data, to improve existing HRA methods. Furthermore, this paper also serves as a demonstration of the value of Bayesian methods to improve the technical basis of HRA.

  1. SPAR Model Structural Efficiencies

    SciTech Connect (OSTI)

    John Schroeder; Dan Henry

    2013-04-01

    The Nuclear Regulatory Commission (NRC) and the Electric Power Research Institute (EPRI) are supporting initiatives aimed at improving the quality of probabilistic risk assessments (PRAs). Included in these initiatives are the resolution of key technical issues that are have been judged to have the most significant influence on the baseline core damage frequency of the NRC’s Standardized Plant Analysis Risk (SPAR) models and licensee PRA models. Previous work addressed issues associated with support system initiating event analysis and loss of off-site power/station blackout analysis. The key technical issues were: • Development of a standard methodology and implementation of support system initiating events • Treatment of loss of offsite power • Development of standard approach for emergency core cooling following containment failure Some of the related issues were not fully resolved. This project continues the effort to resolve outstanding issues. The work scope was intended to include substantial collaboration with EPRI; however, EPRI has had other higher priority initiatives to support. Therefore this project has addressed SPAR modeling issues. The issues addressed are • SPAR model transparency • Common cause failure modeling deficiencies and approaches • Ac and dc modeling deficiencies and approaches • Instrumentation and control system modeling deficiencies and approaches

  2. Science-Based Simulation Model of Human Performance for Human Reliability Analysis

    SciTech Connect (OSTI)

    Dana L. Kelly; Ronald L. Boring; Ali Mosleh; Carol Smidts

    2011-10-01

    Human reliability analysis (HRA), a component of an integrated probabilistic risk assessment (PRA), is the means by which the human contribution to risk is assessed, both qualitatively and quantitatively. However, among the literally dozens of HRA methods that have been developed, most cannot fully model and quantify the types of errors that occurred at Three Mile Island. Furthermore, all of the methods lack a solid empirical basis, relying heavily on expert judgment or empirical results derived in non-reactor domains. Finally, all of the methods are essentially static, and are thus unable to capture the dynamics of an accident in progress. The objective of this work is to begin exploring a dynamic simulation approach to HRA, one whose models have a basis in psychological theories of human performance, and whose quantitative estimates have an empirical basis. This paper highlights a plan to formalize collaboration among the Idaho National Laboratory (INL), the University of Maryland, and The Ohio State University (OSU) to continue development of a simulation model initially formulated at the University of Maryland. Initial work will focus on enhancing the underlying human performance models with the most recent psychological research, and on planning follow-on studies to establish an empirical basis for the model, based on simulator experiments to be carried out at the INL and at the OSU.

  3. Set Equation Transformation System.

    Energy Science and Technology Software Center (OSTI)

    2002-03-22

    Version 00 SETS is used for symbolic manipulation of Boolean equations, particularly the reduction of equations by the application of Boolean identities. It is a flexible and efficient tool for performing probabilistic risk analysis (PRA), vital area analysis, and common cause analysis. The equation manipulation capabilities of SETS can also be used to analyze noncoherent fault trees and determine prime implicants of Boolean functions, to verify circuit design implementation, to determine minimum cost fire protectionmore » requirements for nuclear reactor plants, to obtain solutions to combinatorial optimization problems with Boolean constraints, and to determine the susceptibility of a facility to unauthorized access through nullification of sensors in its protection system. Two auxiliary programs, SEP and FTD, are included. SEP performs the quantitative analysis of reduced Boolean equations (minimal cut sets) produced by SETS. The user can manipulate and evaluate the equations to find the probability of occurrence of any desired event and to produce an importance ranking of the terms and events in an equation. FTD is a fault tree drawing program which uses the proprietary ISSCO DISSPLA graphics software to produce an annotated drawing of a fault tree processed by SETS. The DISSPLA routines are not included.« less

  4. Desert architecture for educational buildings, a case study: A center for training university graduates

    SciTech Connect (OSTI)

    Ebeid, M.

    1996-10-01

    A new program for training graduates in desert development is being implemented by the Desert Development Center (DDC) of the American University in Cairo. The facilities consist of fifty bed/sitting rooms for accommodating 100 students. Each unit consists of two rooms and a bathroom for the use of 4 students; a lecture theater which can house 120 students, with adjoining office for trainers as well as necessary facilities; a general cafeteria which can serve 120--150 persons and an adjoining dining room for teaching staff. The cafeteria building also houses the kitchen; a cold storage area; a laundry room, storerooms, sleeping quarters and services for the labor force of the building complex; a system of solar water heaters; and a special sanitary sewage system for treatment of waste water produced by the building`s activities. When designing and implementing this complex, architectural elements and building philosophy based on the concept of integrating with the environment were considered. Elements included orientation heights and building materials suited to the desert environment, thick walls, outer and inner finishing materials, roofs, malkafs, floors, colors, solar heaters, lighting, green areas, windbreaks, terraces, and furniture. The paper includes a general evaluation of this educational building based on the PRA approach (Participatory Rapid Appraisal) involving those living and working in it. As a result of her position with the project, the author was able to evaluate the original designs, recommend modifications, and evaluate their implementation and fulfillment of the original goals of the projects.

  5. Proof-of-Concept Demonstrations for Computation-Based Human Reliability Analysis. Modeling Operator Performance During Flooding Scenarios

    SciTech Connect (OSTI)

    Joe, Jeffrey Clark; Boring, Ronald Laurids; Herberger, Sarah Elizabeth Marie; Mandelli, Diego; Smith, Curtis Lee

    2015-09-01

    The United States (U.S.) Department of Energy (DOE) Light Water Reactor Sustainability (LWRS) program has the overall objective to help sustain the existing commercial nuclear power plants (NPPs). To accomplish this program objective, there are multiple LWRS “pathways,” or research and development (R&D) focus areas. One LWRS focus area is called the Risk-Informed Safety Margin and Characterization (RISMC) pathway. Initial efforts under this pathway to combine probabilistic and plant multi-physics models to quantify safety margins and support business decisions also included HRA, but in a somewhat simplified manner. HRA experts at Idaho National Laboratory (INL) have been collaborating with other experts to develop a computational HRA approach, called the Human Unimodel for Nuclear Technology to Enhance Reliability (HUNTER), for inclusion into the RISMC framework. The basic premise of this research is to leverage applicable computational techniques, namely simulation and modeling, to develop and then, using RAVEN as a controller, seamlessly integrate virtual operator models (HUNTER) with 1) the dynamic computational MOOSE runtime environment that includes a full-scope plant model, and 2) the RISMC framework PRA models already in use. The HUNTER computational HRA approach is a hybrid approach that leverages past work from cognitive psychology, human performance modeling, and HRA, but it is also a significant departure from existing static and even dynamic HRA methods. This report is divided into five chapters that cover the development of an external flooding event test case and associated statistical modeling considerations.

  6. Generating human reliability estimates using expert judgment. Volume 1. Main report

    SciTech Connect (OSTI)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessment (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 1 of this report provides a brief overview of the background of the project, the procedure for using psychological scaling techniques to generate HEP estimates and conclusions from evaluation of the techniques. Results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. In addition, HEP estimates for 35 tasks related to boiling water reactors (BMRs) were obtained as part of the evaluation. These HEP estimates are also included in the report.

  7. Generating human reliability estimates using expert judgment. Volume 2. Appendices. [PWR; BWR

    SciTech Connect (OSTI)

    Comer, M.K.; Seaver, D.A.; Stillwell, W.G.; Gaddy, C.D.

    1984-11-01

    The US Nuclear Regulatory Commission is conducting a research program to determine the practicality, acceptability, and usefulness of several different methods for obtaining human reliability data and estimates that can be used in nuclear power plant probabilistic risk assessments (PRA). One method, investigated as part of this overall research program, uses expert judgment to generate human error probability (HEP) estimates and associated uncertainty bounds. The project described in this document evaluated two techniques for using expert judgment: paired comparisons and direct numerical estimation. Volume 2 provides detailed procedures for using the techniques, detailed descriptions of the analyses performed to evaluate the techniques, and HEP estimates generated as part of this project. The results of the evaluation indicate that techniques using expert judgment should be given strong consideration for use in developing HEP estimates. Judgments were shown to be consistent and to provide HEP estimates with a good degree of convergent validity. Of the two techniques tested, direct numerical estimation appears to be preferable in terms of ease of application and quality of results.

  8. Regulatory cross-cutting topics for fuel cycle facilities.

    SciTech Connect (OSTI)

    Denman, Matthew R.; Brown, Jason; Goldmann, Andrew Scott; Louie, David

    2013-10-01

    This report overviews crosscutting regulatory topics for nuclear fuel cycle facilities for use in the Fuel Cycle Research&Development Nuclear Fuel Cycle Evaluation and Screening study. In particular, the regulatory infrastructure and analysis capability is assessed for the following topical areas:Fire Regulations (i.e., how applicable are current Nuclear Regulatory Commission (NRC) and/or International Atomic Energy Agency (IAEA) fire regulations to advance fuel cycle facilities)Consequence Assessment (i.e., how applicable are current radionuclide transportation tools to support risk-informed regulations and Level 2 and/or 3 PRA) While not addressed in detail, the following regulatory topic is also discussed:Integrated Security, Safeguard and Safety Requirement (i.e., how applicable are current Nuclear Regulatory Commission (NRC) regulations to future fuel cycle facilities which will likely be required to balance the sometimes conflicting Material Accountability, Security, and Safety requirements.)

  9. Multidisciplinary framework for human reliability analysis with an application to errors of commission and dependencies

    SciTech Connect (OSTI)

    Barriere, M.T.; Luckas, W.J.; Wreathall, J.; Cooper, S.E.; Bley, D.C.; Ramey-Smith, A.

    1995-08-01

    Since the early 1970s, human reliability analysis (HRA) has been considered to be an integral part of probabilistic risk assessments (PRAs). Nuclear power plant (NPP) events, from Three Mile Island through the mid-1980s, showed the importance of human performance to NPP risk. Recent events demonstrate that human performance continues to be a dominant source of risk. In light of these observations, the current limitations of existing HRA approaches become apparent when the role of humans is examined explicitly in the context of real NPP events. The development of new or improved HRA methodologies to more realistically represent human performance is recognized by the Nuclear Regulatory Commission (NRC) as a necessary means to increase the utility of PRAS. To accomplish this objective, an Improved HRA Project, sponsored by the NRC`s Office of Nuclear Regulatory Research (RES), was initiated in late February, 1992, at Brookhaven National Laboratory (BNL) to develop an improved method for HRA that more realistically assesses the human contribution to plant risk and can be fully integrated with PRA. This report describes the research efforts including the development of a multidisciplinary HRA framework, the characterization and representation of errors of commission, and an approach for addressing human dependencies. The implications of the research and necessary requirements for further development also are discussed.

  10. Analysis of core damage frequency: Peach Bottom, Unit 2 internal events appendices

    SciTech Connect (OSTI)

    Kolaczkowski, A.M.; Cramond, W.R.; Sype, T.T.; Maloney, K.J.; Wheeler, T.A.; Daniel, S.L.; Sandia National Labs., Albuquerque, NM )

    1989-08-01

    This document contains the appendices for the accident sequence analysis of internally initiated events for the Peach Bottom, Unit 2 Nuclear Power Plant. This is one of the five plant analyses conducted as part of the NUREG-1150 effort for the Nuclear Regulatory Commission. The work performed and described here is an extensive reanalysis of that published in October 1986 as NUREG/CR-4550, Volume 4. It addresses comments from numerous reviewers and significant changes to the plant systems and procedures made since the first report. The uncertainty analysis and presentation of results are also much improved, and considerable effort was expended on an improved analysis of loss of offsite power. The content and detail of this report is directed toward PRA practitioners who need to know how the work was done and the details for use in further studies. The mean core damage frequency is 4.5E-6 with 5% and 95% uncertainty bounds of 3.5E-7 and 1.3E-5, respectively. Station blackout type accidents (loss of all ac power) contributed about 46% of the core damage frequency with Anticipated Transient Without Scram (ATWS) accidents contributing another 42%. The numerical results are driven by loss of offsite power, transients with the power conversion system initially available operator errors, and mechanical failure to scram. 13 refs., 345 figs., 171 tabs.