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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. 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.

  3. INCORPORATING DYNAMIC 3D SIMULATION INTO PRA

    SciTech Connect (OSTI)

    Steven R Prescott; Curtis Smith

    2011-07-01

    Through continued advancement in computational resources, development that was previously done by trial and error production is now performed through computer simulation. These virtual physical representations have the potential to provide accurate and valid modeling results and are being used in many different technical fields. Risk assessment now has the opportunity to use 3D simulation to improve analysis results and insights, especially for external event analysis. By using simulations, the modeler only has to determine the likelihood of an event without having to also predict the results of that event. The 3D simulation automatically determines not only the outcome of the event, but when those failures occur. How can we effectively incorporate 3D simulation into traditional PRA? Most PRA plant modeling is made up of components with different failure modes, probabilities, and rates. Typically, these components are grouped into various systems and then are modeled together (in different combinations) as a “system” with logic structures to form fault trees. Applicable fault trees are combined through scenarios, typically represented by event tree models. Though this method gives us failure results for a given model, it has limitations when it comes to time-based dependencies or dependencies that are coupled to physical processes which may themselves be space- or time-dependent. Since, failures from a 3D simulation are naturally time related, they should be used in that manner. In our simulation approach, traditional static models are converted into an equivalent state diagram representation with start states, probabilistic driven movements between states and terminal states. As the state model is run repeatedly, it converges to the same results as the PRA model in cases where time-related factors are not important. In cases where timing considerations are important (e.g., when events are dependent upon each other), then the simulation approach will typically provide superior results and insights. We also couple the state model with the dynamic 3D simulation analysis representing events (such as flooding) to determine which (if any) components fail. Not only does the simulation take into account any failed items from the state model, but any failures caused by the simulation are incorporated back into the state model and factored into the overall results. Using this method we incorporate accurate 3D simulation results, eliminate static-based PRA issues, and have time ordered failure information.

  4. 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.

  5. 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.

  6. An evaluation of the reliability and usefulness of external-initiator PRA (probabilistic risk analysis) methodologies

    SciTech Connect (OSTI)

    Budnitz, R.J.; Lambert, H.E. (Future Resources Associates, Inc., Berkeley, CA (USA))

    1990-01-01

    The discipline of probabilistic risk analysis (PRA) has become so mature in recent years that it is now being used routinely to assist decision-making throughout the nuclear industry. This includes decision-making that affects design, construction, operation, maintenance, and regulation. Unfortunately, not all sub-areas within the larger discipline of PRA are equally mature,'' and therefore the many different types of engineering insights from PRA are not all equally reliable. 93 refs., 4 figs., 1 tab.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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 events and eighteen SBO sequences (versus fifty-four SBO sequences in the original SBO model). The estimated SBO sequence conditional probabilities from the original SBO model were integrated to the corresponding sequences in the simplified SBO event tree. These results were then compared with the simulation run results.

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1988-01-01

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

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

    SciTech Connect (OSTI)

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

    1993-12-31

    Traditionally, probabilistic risk assessments (PRA) of severe accidents in nuclear power plants have considered initiating events potentially occurring only during full power operation. Some previous screening analysis that were performed for other modes of operation suggested that risks during those modes were small relative to full power operation. However, more recent studies and operational experience have implied that accidents during low power and shutdown could be significant contributors to risk. The objective of this paper is to present the approach utilized in the level-1 PRA for the Surry plant, and discuss the results obtained. A comparison of the results with those of other shutdown studies is provided. Relevant safety issues such as plant and hardware configurations, operator training, and instrumentation and control is discussed.

  16. 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.

  17. Limitations imposed on fire PRA methods as the result of incomplete and uncertain fire event data.

    SciTech Connect (OSTI)

    Nowlen, Steven Patrick; Hyslop, J. S. (U.S. Nuclear Regulatory Commission, Washington, DC)

    2010-04-01

    Fire probabilistic risk assessment (PRA) methods utilize data and insights gained from actual fire events in a variety of ways. For example, fire occurrence frequencies, manual fire fighting effectiveness and timing, and the distribution of fire events by fire source and plant location are all based directly on the historical experience base. Other factors are either derived indirectly or supported qualitatively based on insights from the event data. These factors include the general nature and intensity of plant fires, insights into operator performance, and insights into fire growth and damage behaviors. This paper will discuss the potential methodology improvements that could be realized if more complete fire event reporting information were available. Areas that could benefit from more complete event reporting that will be discussed in the paper include fire event frequency analysis, analysis of fire detection and suppression system performance including incipient detection systems, analysis of manual fire fighting performance, treatment of fire growth from incipient stages to fully-involved fires, operator response to fire events, the impact of smoke on plant operations and equipment, and the impact of fire-induced cable failures on plant electrical circuits.

  18. Optimal Estimation of Dynamically Evolving Di usivities Kurt S. Riedel

    E-Print Network [OSTI]

    equations to account for model error [15]. Researchers attempt to model the e#11;ect of microscopic turbulence in plasmas and uids with anomalous di#11;usion coeÆcients. These e#11;ective equations for uid ow are only an approximation of the actual evolution equations, and in many cases the model error

  19. ccsd00002004, Di usivity induced by vortex-like coherent

    E-Print Network [OSTI]

    in Reversed Field Pinch plasmas M. Spolaore x, V. Antoni, E. Spada, R. Cavazzana, E. Martines, G. Regnoli z, G Laboratory, Royal Institute of Technology, SE10044, Stockholm, Sweden Abstract. Coherent structures emerging of two Reversed Field Pinch experiments, RFX (Padua) and Extrap-T2R (Stockholm). Measurements have been

  20. USING STANDARD PRA S

    Office of Scientific and Technical Information (OSTI)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of NaturalDukeWakefield MunicipalTechnicalInformation4563 LLNL Small-scale Friction Sensitivity (BAM) Test .ll11Directors

  1. 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

  2. Shear wall ultimate drift limits for PRA applications

    SciTech Connect (OSTI)

    Duffey, T.A. [New Mexico Highlands Univ., Las Vegas, NM (United States); Farrar, C.R.; Goldman, A. [Los Alamos National Lab., NM (United States)

    1995-03-01

    Drift limits for reinforced concrete shear walls are investigated by reviewing the technical literature for appropriate experimental data. Based on the geometry of actual nuclear power plant structures (exclusive of containments) and concerns regarding their response during seismic loading, data are obtained from pertinent references where the wall aspect ratio is less than or equal to approximately 1, and for which the loading is cyclic. Lateral deflections at ultimate load, and at points in the softening region beyond ultimate, are obtained and converted to drift information. The statistical nature of the data is also investigated. These data are shown to be lognormally distributed, and an analysis of variance is performed. The use of these statistics to estimate Probability of Failure for a shear wall structure is illustrated.

  3. 1 INTRODUCTION Probabilistic risk (or safety) assessments (PRA) pro-

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    reliability analyses. Finally, a case study in- volving a nuclear reactor is presented in Section 3. Dynamic for managing risks linked to engineering systems, notably in nuclear power plants, aerospace, and chemical of dynamic reliability was established under the name of Con- tinuous Event Tree (CET) theory, (Devooght

  4. 01Aniversariantes do dia: 01 ALFREDO BUENO ASSISTENTE EM ADMINISTRACAO PRA/RU

    E-Print Network [OSTI]

    Paraná, Universidade Federal do

    ZAVORNE TECNICO EM RADIOLOGIA HC 02 JOSE GERALDO AUERSWALD CALOMENO PROFESSOR DO MAGISTERIO SUPERIOR BL

  5. A PRA~ATIC CONCEPT OF THEME AND RHEME FOR MACHINE Christa Hauenschild

    E-Print Network [OSTI]

    , Universitat Konstanz, Postfach 5560, D-7750 Konst anz The concept of theme and rheme we want to propose

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

    Office of Environmental Management (EM)

    National Optimization Strategy to meet goals * Goal: Expand optimization throughout pipeline * Goal: Increase number of sites optimized * Goal: Expand optimization resource...

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

    Energy Savers [EERE]

    (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...

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

    Office of Environmental Management (EM)

    12:40 pm Upcoming Activities (Ming Zhu, DOE EM) Webinar Instructions Performance & Risk Assessment Community of Practice Thursday, October 16, 2014 11:00 am - 12:40 pm, Eastern...

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

    Office of Environmental Management (EM)

    2:55 - 3:00 pm Closing (Dr. Ming Zhu, DOE EM) Webinar Instructions Performance & Risk Assessment Community of Practice Webinar Thursday, November 12, 2015 1:00 pm | Eastern Time...

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

    Office of Environmental Management (EM)

    2:55 - 3:00 pm Closing (Dr. Ming Zhu, DOE EM) Webinar Instructions Performance & Risk Assessment Community of Practice Webinar Tuesday, November 10, 2015 1:00 pm | Eastern...

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

    Office of Environmental Management (EM)

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

  12. 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)

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

    Office of Environmental Management (EM)

    Dept 24 Jennifer Heimberg National Academy of Sciences 25 Doug Hildebrand DOE RL 26 Mary Hill University of Kansas 27 Britt Jacobson Nevada State DEP 28 Wayne Johnson PNNL 29...

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

    Office of Environmental Management (EM)

    May 20, 2015, Wednesday, 1:30 am to 3:30 pm EDT Agenda 1:30 - 1:35 am Introduction (Ming Zhu, DOE EM) 1:35 am - 3:20 pm Presentation - Overview of Proposed Guidance for Conducting...

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

    Office of Environmental Management (EM)

    June 3 (Tuesday), 2014, 11:30 am to 1:15 pm EDT Agenda 11:30 - 11:35 Introductions (Ming Zhu, DOE EM) 11:35 - 12:35 Presentation - Features, Events, and Processes (FEPs):...

  16. 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.

  17. Microsoft Word - PRA CoP Techncial Exchange Draft Agenda 2015-11-02

    Office of Environmental Management (EM)

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantity of Natural GasAdjustments (BillionProvedTravelInformation Resources»Jim1 ENVIRONMENTAL MANAGEMENTJuly 9,0 of 4OAK RIDGE

  18. PoS(PRA2009)028 The ATLAS Survey of the CDFS and ELAIS-S1 Fields

    E-Print Network [OSTI]

    Norris, Ray

    has been designed to use ASKAP (Australian Square Kilometre Array Pathfinder). Panoramic Radio Survey) project surveyed a total 7 square degrees down to 30 µJy rms at 1.4 GHz and is the largest surveyed a total 7 square degrees down to 30 µJy rms at 1.4 GHz and is the largest sensitive radio survey

  19. 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...

  20. Transient thermal behaviour of a solid oxide fuel cell Moussa Chnani, Marie-Ccile Pra, Raynal Glises, Jean Marie Kauffmann and

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    and Electrochemical modelling. 1- Introduction The solid oxide fuel cell (SOFC) is a promising technologyTransient thermal behaviour of a solid oxide fuel cell Moussa Chnani, Marie-Cécile Péra, Raynal provided by HTceramix. Keywords: Solid oxide fuel cell; Transient thermal modelling; Fluidic

  1. Genomic analysis of organismal complexity in the multicellular green alga Volvox carteri

    E-Print Network [OSTI]

    Prochnik, Simon E.

    2011-01-01

    Ddi Tth Pra Pso Neurospora crassa Prochlorococcus marinusPra, Phytophthora ramorum; Pso, Phytophthora sojae; Ncr,Au9.Cre12.g517400 Cme Ddi Tth Pra Pso Ncr Ath Hsa Cel Ota

  2. Pedro de Castañeda y Nájera, Relación de la Jornada de Cíbola: acotaciones gramaticales y léxicas

    E-Print Network [OSTI]

    Craddock, Jerry R.

    2010-01-01

    de su opinion pusiesen en pra- | tica la buelta de la nueuamanuscrito se lee “pra- | tica” (CyN 130r5-6) por plática ‘

  3. The Magnetic Nature of Solar Flares E.R. Priest + and T.G. Forbes

    E-Print Network [OSTI]

    Priest, Eric

    , Oceans and Space, University of New Hampshire, Durham NH 03824, USA Abstract The main challenge turbulence, acceleration by direct electric #12;elds 1 #12; at the reconnection site, or di#11;usive shock an active region is much lower and so the eruptive speeds and electric #12;elds are much smaller, so

  4. * Corresponding author. Present address: Department of Oceanography, University of Hawai'i, MSB 610, 1000 Pope Road, Honolulu, HI 96822, USA. Tel.: (808) 956-7625; fax: (808) 956-9516.

    E-Print Network [OSTI]

    Hinch, Scott G.

    contaminants, such as anthropogenically produced CO , hydrocarbons and heavy metals (Wageman and Muir, 1984 estimate. Be derived estimates of vertical eddy di!usivity into summer surface waters ranged from 0.5 to 1 (Brooks, 1991). This semi-enclosed water mass is bordered by a heavily populated coastline and supports

  5. 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...

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

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

    Updates on the Performance and Risk Assessment Community of Practice (P&RA CoP) Status Updates on the Performance and Risk Assessment Community of Practice (P&RA CoP) Ming Zhu,...

  7. Microsoft PowerPoint - 0 Ming Zhu

    Office of Environmental Management (EM)

    Status Updates on the Performance and Risk Assessment Community of Practice (P&RA CoP) P&RA CoP Technical Exchange Meeting Las Vegas, NV December 11-12, 2014 Ming Zhu, Ph.D., PE,...

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

    Energy Savers [EERE]

    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...

  9. 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.

  10. CV

    E-Print Network [OSTI]

    2015-10-24

    Pra lat, editors, Algorithms and Models for the Web Graph, volume 8305 of Lecture Notes in. Computer Science, pages 68—79. Springer International Publishing ...

  11. 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...

  12. Aldrich -N32407 Sigma-Aldrich Corporation

    E-Print Network [OSTI]

    Choi, Kyu Yong

    in accordance with good industrial hygiene and safety pra to the amount and concentration of the dangerous substance at the work place. Hygiene measures Handle

  13. Associate Directorate for Environmental Programs Update January...

    Broader source: Energy.gov (indexed) [DOE]

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

  14. 2008-10 "Independent Review of the MDA G CME Report by DOE Office...

    Broader source: Energy.gov (indexed) [DOE]

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

  15. Confidentiality Agreement between the Nuclear Decommissioning...

    Office of Environmental Management (EM)

    Decommissioning (D&D) P&RA Community of Practice Facility Engineering Soil & Groundwater Sustainability Program Management Safety Security Quality Assurance Budget & Performance...

  16. December 12, 2013 Webinar - The Use of Graded Approach in Hanford...

    Office of Environmental Management (EM)

    Protection Application of Model Abstraction Techniques to Simulate Transport in Soils (NUREGCR-7026) Agenda - 12122013 - P&RA CoP Webinar Presentation - Hanford Site...

  17. August 18, 2015 Webinar - Probabilistic Analysis of Inadvertent...

    Broader source: Energy.gov (indexed) [DOE]

    Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - August 18, 2015 - Probabilistic Analysis of Inadvertent Intrusion and the International Atomic Energy...

  18. 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...

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

    Energy Savers [EERE]

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

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

    Office of Environmental Management (EM)

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

  1. Introduction to DOE Order 435.1 Low Level Radioactive Waste Disposal...

    Energy Savers [EERE]

    Disposal Requirements More Documents & Publications Status Updates on the Performance and Risk Assessment Community of Practice (P&RA CoP) LFRG Program Management Plan LFRG Charter...

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

    Office of Environmental Management (EM)

    Topics for Interagency Performance & Risk Assessment Community of Practice (P&RA CoP) Discussion 11122015 Webinars being conducted or planned for 20152016: * Guidance for...

  3. Information Collection Management | Department of Energy

    Energy Savers [EERE]

    Information Collection RequestsPRA (PDF) DOE Order 200.2 Information Collection Management Program - To set forth the Department of Energy (DOE) requirements and...

  4. The High Cost of Compromise: Tobacco Industry Political Influence and Tobacco Control Policy in Virginia, 1977-2009

    E-Print Network [OSTI]

    Kierstein, Alex JD; Barnes, Richard L. JD; Glantz, Stanton A. PhD

    2010-01-01

    E, Foreman L, Pra MD. The Post-Buyout Experience: Peanut andhave included a “quota buyout” to compensate existing quotamanufacturers over a quota buyout to garner growers’ support

  5. Procedures for Obtaining OMB Clearance to Conduct a Survey

    SciTech Connect (OSTI)

    None

    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.

  6. Technical Session IV Talks | U.S. DOE Office of Science (SC)

    Office of Science (SC) Website

    AFDC Printable Version Share this resource Send a link to EERE: Alternative Fuels Data Center Home Page to someone by E-mail Share EERE: Alternative Fuels Data Center Home Page on Facebook Tweet about EERE: Alternative Fuels Data Center Home Page on Twitter Bookmark EERE: Alternative Fuels Data Center Homesum_a_epg0_fpd_mmcf_m.xls" ,"Available from WebQuantityBonneville Power AdministrationRobust,Field-effectWorkingLosThe23-24, 2011Science (SC) RedefiningSign Up| U.S.IV Talks

  7. Pests and Diseases 1 Exotic longhorn beetle

    E-Print Network [OSTI]

    the risks posed by pests from the former Soviet Union. Meetings in Helsinki (Finland), Perm (Russia. This approach has the benefit of including expertise from both the importing (the PRA area) and exporting exporting and importing countries alike. In relation to PRA technology, scientists from both Branches have

  8. Atmos. Chem. Phys., 9, 74617479, 2009 www.atmos-chem-phys.net/9/7461/2009/

    E-Print Network [OSTI]

    Meskhidze, Nicholas

    Sciences, Institute for Terrestrial and Planetary Atmospheres, Stony Brook University, Stony Brook, NY, USA scenario over a period of five days. By cou- pling the P¨oschl-Rudich-Ammann (PRA) kinetic framework. The flux-based PRA formula- tion takes into account changes in the uptake kinetics due to changes

  9. 1 Copyright 2011 by ASME Proceedings of the ASME 2011 Power Conference

    E-Print Network [OSTI]

    Rubloff, Gary W.

    . College Park, MD, USA Mohammad Modarres University of Maryland Department of Mechanical Eng. College Park, MD, USA Aris Christou University of Maryland Department of Mechanical Eng. College Park, MD, USA [3] that led to the advent of PRA in the nuclear industry. Over the years, PRA has grown

  10. Two sixteenth century chroniclers and the Indian policy of the Spanish state 

    E-Print Network [OSTI]

    Huffman, Sarah Phillips

    1977-01-01

    Castro, De la edad conf lictiva (Madrid: Taurus Ediciones, 1972), and La realidad histdrica de ~Es aPra (Mexico: Editorial Porrua, 1973+ 19 Liss, p. 34. 21 CHAPTER III SPANISH INDIAN POLICY: THEORIES AND PHILOSOPHIES Lewis Hanke, an eminent...

  11. Criteria for assessing the quality of nuclear probabilistic risk assessments

    E-Print Network [OSTI]

    Zhu, Yingli, 1976-

    2004-01-01

    The final outcome of a nuclear Probabilistic Risk Assessment (PRA) is generally inaccurate and imprecise. This is primarily because not all risk contributors are addressed in the analysis, and there are state-of-knowledge ...

  12. Hanford Site Waste Management Area C Performance Assessment

    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.

  13. Probabilistic Risk Assessment for dairy waste management systems 

    E-Print Network [OSTI]

    Leigh, Edward Marshall

    1993-01-01

    Probabilistic Risk Assessment (PRA) techniques were used to evaluate the risk of contamination of surface and ground water with wastewater from an open lot dairy in Erath County, Texas. The dairy supported a complex waste management system...

  14. Quality Assurance for Performance Assessment Modeling

    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. 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.

  16. 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.

  17. A Bayesian approach to integrate temporal data into probabilistic risk analysis of monitored NAPL remediation

    E-Print Network [OSTI]

    Bolster, Diogo

    A Bayesian approach to integrate temporal data into probabilistic risk analysis of monitored NAPL quantifying risks associated with the failure of such efforts. We conduct a probabilistic risk analysis (PRA

  18. 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.

  19. 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.

  20. Channeling in purine biosynthesis : efforts to detect interactions between PurF and PurD and characterization of the FGAR-AT complex

    E-Print Network [OSTI]

    Hoskins, Aaron A. (Aaron Andrew)

    2006-01-01

    Purine biosynthesis has been used as a paradigm for the study of metabolism of unstable molecules. Both phosphoribosylamine (PRA) and N5-carboxyaminoimidazole ribonucleotide (N5-CAIR) have estimated half-lives in vivo of ...

  1. vec l'hritage de deux grands foyers de peu-

    E-Print Network [OSTI]

    Boyer, Edmond

    140 km2 , soit une densité moyenne de 658 habitants/km2 . Les deux États de l'extrême sud, le Kerala démographiques différenciées. Les quatre États du sud - Kerala, Tamil Nadu, Andhra Pra- desh et Karnataka ­ ont

  2. Satellite System Safety Analysis Using STPA

    E-Print Network [OSTI]

    Dunn, Nicholas Connor

    2013-01-01

    Traditional hazard analysis techniques based on failure models of accident causality, such as the probabilistic risk assessment (PRA) method currently used at NASA, are inadequate for analyzing safety at the system level. ...

  3. 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.

  4. 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.

  5. November 10, 2015 Webinar - Congressionally Mandated Review of...

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

    Review of the Use Of Risk-Informed Management in the DOE EM Program Performance & RIsk Assessment (P&RA) Community of Practice (CoP) Webinar - November 10, 2015 Webinar -...

  6. Assessing the performance of human-automation collaborative planning systems

    E-Print Network [OSTI]

    Ryan, Jason C. (Jason Christopher)

    2011-01-01

    Planning and Resource Allocation (P/RA) Human Supervisory Control (HSC) systems utilize the capabilities of both human operators and automated planning algorithms to schedule tasks for complex systems. In these systems, ...

  7. 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.

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

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

    Advanced Radiation Medicine Dates Scheduled Register Online April 11-15, 2016 August 15-19, 2016 Fee: 275 Maximum enrollment: 28 30 hours AMA PRA Category 1 Credits(tm) This...

  9. A framework for dynamic safety and risk management modeling in complex engineering systems

    E-Print Network [OSTI]

    Dulac, Nicolas, 1978-

    2007-01-01

    Almost all traditional hazard analysis or risk assessment techniques, such as failure modes and effect analysis (FMEA), fault tree analysis (FTA), and probabilistic risk analysis (PRA) rely on a chain-of-event paradigm of ...

  10. Polar rotation angle identifies elliptic islands in unsteady dynamical systems

    E-Print Network [OSTI]

    Mohammad Farazmand; George Haller

    2015-03-20

    We propose rotation inferred from the polar decomposition of the flow gradient as a diagnostic for elliptic (or vortex-type) invariant regions in non-autonomous dynamical systems. We consider here two- and three-dimensional systems, in which polar rotation can be characterized by a single angle. For this polar rotation angle (PRA), we derive explicit formulas using the singular values and vectors of the flow gradient. We find that closed level sets of the PRA reveal elliptic islands in great detail, and singular level sets of the PRA uncover centers of such islands. Both features turn out to be objective (frame-invariant) for two-dimensional systems. We illustrate the diagnostic power of PRA for elliptic structures on several examples.

  11. Polar rotation angle identifies elliptic islands in unsteady dynamical systems

    E-Print Network [OSTI]

    Farazmand, Mohammad

    2015-01-01

    We propose rotation inferred from the polar decomposition of the flow gradient as a diagnostic for elliptic (or vortex-type) invariant regions in non-autonomous dynamical systems. We consider here two- and three-dimensional systems, in which polar rotation can be characterized by a single angle. For this polar rotation angle (PRA), we derive explicit formulas using the singular values and vectors of the flow gradient. We find that closed level sets of the PRA reveal elliptic islands in great detail, and singular level sets of the PRA uncover centers of such islands. Both features turn out to be objective (frame-invariant) for two-dimensional systems. We illustrate the diagnostic power of PRA for elliptic structures on several examples.

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

    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.

  13. 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.

  14. Toward resilient communities: A performance-based engineering framework for design and evaluation of the built environment

    E-Print Network [OSTI]

    Mieler, Michael

    2012-01-01

    National Standards Institute (ANSI) and American Nuclearevents PRA methodology (ANSI/ANS-58.21-2003). AmericanAboveground Piping ASME/ANSI B31.4 none ASCE TCLEE 1984

  15. The use for frequency-consequence curves in future reactor licensing

    E-Print Network [OSTI]

    Debesse, Laurène

    2007-01-01

    The licensing of nuclear power plants has focused until now on Light Water Reactors and has not incorporated systematically insights and benefits from Probabilistic Risk Assessment (PRA). With the goal of making the licensing ...

  16. Uncertainty and sensitivity analysis of a fire-induced accident scenario involving binary variables and mechanistic codes

    E-Print Network [OSTI]

    Minton, Mark A. (Mark Aaron)

    2010-01-01

    In response to the transition by the United States Nuclear Regulatory Commission (NRC) to a risk-informed, performance-based fire protection rulemaking standard, Fire Probabilistic Risk Assessment (PRA) methods have been ...

  17. 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.

  18. ICONE-4: Proceedings. Volume 3: Safety and reliability

    SciTech Connect (OSTI)

    Rao, A.S. [ed.] [General Electric Nuclear Engineering, San Jose, CA (United States); Duffey, R.B. [ed.] [Brookhaven National Lab., Upton, NY (United States); Elias, D. [ed.] [Commonwealth Edison, Downers Grove, IL (United States)

    1996-07-01

    The proceedings of this conference are divided into five volumes. This volume is divided into the following sections: operation and maintenance; accident simulation and analysis; stability analysis; Korea Project -- safety analysis and radiation release; severe accident analysis; severe accident features; severe accident management; emerging risk-based PRA applications; PSA/PRA applications; PSA -- applications and procedures; PSA -- procedure and policy; dynamic PSA applications; and PSA -- policy and miscellaneous. Separate abstract were prepared for all papers in this volume.

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

    SciTech Connect (OSTI)

    Moore, A. S. [Directorate Science and Technology, AWE Aldermaston, Reading (United Kingdom); Cooper, A. B.R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Schneider, M. B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); MacLaren, S. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Graham, P. [Directorate Science and Technology, AWE Aldermaston, Reading (United Kingdom); Lu, K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Seugling, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Satcher, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Klingmann, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Comley, A. J. [Directorate Science and Technology, AWE Aldermaston, Reading (United Kingdom); Marrs, R. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); May, M. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Widmann, K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Glendinning, G. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Castor, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Sain, J. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Back, C. A. [General Atomics, San Diego, CA (United States); Hund, J. [General Atomics, San Diego, CA (United States); Baker, K. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Hsing, W. W. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Foster, J. [Directorate Science and Technology, AWE Aldermaston, Reading (United Kingdom); Young, B. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Young, P. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States)

    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#11;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

  20. 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.

  1. 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.

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1987-01-01

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

  3. 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 system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). This reference guide will introduce the SAPHIRE Version 8.0 software. A brief discussion of the purpose and history of the software is included along with general information such as installation instructions, starting and stopping the program, and some pointers on how to get around inside the program. Next, database concepts and structure are discussed. Following that discussion are nine sections, one for each of the menu options on the SAPHIRE main menu, wherein the purpose and general capabilities for each option are furnished. Next, the capabilities and limitations of the software are provided.

  4. 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.

  5. Updated Risk Analysis of the LHC Cryogenic and Helium Distribution System

    E-Print Network [OSTI]

    Chorowski, M; Tavian, L

    2013-01-01

    The Preliminary Risk Analysis (PRA) of the Large Hadron Collider cryogenic system, performed in 1998, was aimed at the identification of all the risks for personnel, equipment or environment caused by the failures that might accidentally occur in any phase of the machine operation, and that could not be eliminated by design. The risk analysis was performed during a design and an early construction phase of the machine, so after the collider commissioning and consolidation experience, especially due to the 080919 incident in the LHC sector 3-4, PRA had to be revised and updated. The paper discusses the criterions of cryogenic failures categorization taking into account their occurrence and severity.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Our institut Internationa

    E-Print Network [OSTI]

    of Higher E ly transpare social equity rganisation a nd training in o is to partic t studying th eration portunities o cademic staff e of good pra e then, more ng also one o that framew very strict qu so that its Ac

  12. FROM POTS TO VDSL -TECHNICAL ASPECTS OF A REASONABLE MIGRATION STRATEGY

    E-Print Network [OSTI]

    Henkel, Werner

    FROM POTS TO VDSL - TECHNICAL ASPECTS OF A REASONABLE MIGRATION STRATEGY Günther Komp, Thomas. 2. Migration from narrowband to broadband services The network which appears to be well suited (ISDN). High bit rate transmission systems like Primary Rate Access (PRA) and subscriber

  13. A decision-support scheme for mapping endangered areas in pest risk analysis*

    E-Print Network [OSTI]

    Kratochvíl, Lukas

    A decision-support scheme for mapping endangered areas in pest risk analysis* R. H. A. Baker1, J to the EPPO DSS for pest risk analysis (PRA) (http:// capra.eppo.org/deliverables; http Systems Analysis, Crop & Weed Ecology Group, P.O. Box 430, 6700 AK Wageningen (The Netherlands) 13The Bio

  14. Bio 320 Ecology Fall 2010 Problem Set 2 1. A population of 600 individuals experiences spatial heterogeneity in its discrete-time

    E-Print Network [OSTI]

    Caraco, Thomas

    were of 2 types. The annual reproductive rate A takes 2 values, with differing probabilities: Pr[A = 2 in a population with non-overlapping generations is 8. Otherwise, the mean number of individuals produced per according to a logistic model where: dN/dt = 0.2 N ­ 0.0002 N2 What equation describes the individual

  15. Ethan Burns (UNH) Abstraction in Multicore Heuristic Search 1 / 31 Parallel Best-First Search: The Role of Abstraction

    E-Print Network [OSTI]

    Ruml, Wheeler

    Ethan Burns (UNH) Abstraction in Multicore Heuristic Search ­ 1 / 31 Parallel Best-First Search: The Role of Abstraction Ethan Burns1, Sofia Lemons1, Wheeler Ruml1 and Rong Zhou2 1 2 [Many thanks to NSF s Best-first Search s Parallel Search PRA* PBNF Optimal Search Suboptimal Search Conclusion Ethan Burns

  16. Preparation and characterization of porous silica xerogel film for low dielectric application

    E-Print Network [OSTI]

    Jo, Moon-Ho

    microelectronics precursors [2]. In particular, one of the porous SiO2 gels, aerogels, has extremely high por aerogel can be applied to IMD [3­5]. In our previous work, we obtained SiO2 aerogel thin film with good, an ambient drying method for the preparation of SiO2 aerogel film was studied and recently reported by Pra

  17. Universitt Project EURAT

    E-Print Network [OSTI]

    Heermann, Dieter W.

    of individual people in far less time today than at the conclusion of the Human Genome Project in 2003 (Collins of Whole Human Genome Sequencing" Position Paper Cornerstones for an ethiCally and legally informed Pra 60 61 63 65 67 69 70 71 73 88 96 100 #12;4 5Ethical and Legal Aspects of Whole Human Genome

  18. Teaching Materials! 1. PROGRAMS OF STUDY ! ! ! ! ! ! ! ! !

    E-Print Network [OSTI]

    Burg, Theresa

    . ASSESSMENT MATERIAL! ! ! ! ! ! The Drama Resource Manuals (listed in Step 2) contain select rubricsDRAMA ! Teaching Materials! !!! ! 1. PROGRAMS OF STUDY ! ! ! ! ! ! ! ! ! !Fine Arts Program for the Theater 3rd ed.: A Handbook of Teaching an Directing Techniques T 792.028 Pra Gr. 10-12 A Practical

  19. PORTFOLIO RISK ASSESSMENT OF SA WATER'S LARGE DAMS by David S. Bowles1

    E-Print Network [OSTI]

    Bowles, David S.

    PORTFOLIO RISK ASSESSMENT OF SA WATER'S LARGE DAMS by David S. Bowles1 , Andrew M. Parsons2 , Loren R. Anderson3 and Terry F. Glover4 ABSTRACT This paper summarises the Portfolio Risk Assessment (PRA a reconnaissance-level engineering assessment and risk assessment. These assessments were performed for floods

  20. Kheshbn No. 102- Fall 1983 - Journal

    E-Print Network [OSTI]

    1983-01-01

    oVixtfü ,ix»-iypìx apy* UN inox ,*ip *iyw Viap m UN .a .K ,mrax yüyösV u t flö asra py Inox,, puya Tra» tü'Ua D3" ? wapaisya ,opt •nx^pra x u px inox ut px «pia ,ptes pp ps yaa-

  1. September 7 -9, 2007 5th annual rocky mountaIn GerIatrIc conference

    E-Print Network [OSTI]

    Tipple, Brett

    September 7 - 9, 2007 5th annual rocky mountaIn GerIatrIc conference keyStone conference center keyStone AAFP, and AMA Category 1 CME credit for the PRA from organizations accredited by ACCME. Keystone, Colorado is just 90 minutes (105 miles) from Denver. How to get to Keystone: From Denver's International

  2. Neo-Latin News 

    E-Print Network [OSTI]

    Kallendorf, Craig et al

    2009-01-01

    parientes natales que las mujeres ya sufren debido a las pra´cticas de parentesco Kazajo que enfatiza el linaje patrilineal, la exogamia basada en clan, y el matrimonio patrilocal. [ge´nero, migracio´n, parentesco, transnacional, Asia Central] I n 2008, we...

  3. 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 system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with 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. Next, the capabilities and limitations of the software are provided.

  4. 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 system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for ansforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with 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. Next, the capabilities and limitations of the software are provided.

  5. 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.

  6. 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.

  7. 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).

  8. HTGR fuel element structural design considerations

    SciTech Connect (OSTI)

    Alloway, R.; Gorholt, W.; Ho, F.; Vollman, R.; Yu, H.

    1986-09-01

    The structural design of the large HTGR prismatic core fuel elements involve the interaction of four engineering disciplines: nuclear physics, thermo-hydraulics, structural and material science. Fuel element stress analysis techniques and the development of structural criteria are discussed in the context of an overview of the entire design process. The core of the proposed 2240 MW(t) HTGR is described as an example where the design process was used. Probabalistic stress analysis techniques coupled with probabalistic risk analysis (PRA) to develop structural criteria to account for uncertainty are described. The PRA provides a means for ensuring that the proposed structural criteria are consistent with plant investment and safety risk goals. The evaluation of cracked fuel elements removed from the Fort St. Vrain reactor in the USA is discussed in the context of stress analysis uncertainty and structural criteria development.

  9. 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.

  10. 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.

  11. Observation of sub-Doppler absorption in the /Lambda-type three-level Doppler-broadened cesium system

    E-Print Network [OSTI]

    Junmin Wang; Yanhua Wang; Shubin Yan; Tao Liu; Tiancai Zhang

    2003-12-31

    Thanks to the atomic coherence in coupling laser driven atomic system, sub-Doppler absorption has been observed in Doppler-broadened cesium vapor cell via the /Lambda-type three-level scheme. The linewidth of the sub-Doppler absorption peak become narrower while the frequency detuning of coupling laser increases. The results are in agreement with the theoretical prediction by G. Vemuri et al.[PRA,Vol.53(1996) p.2842].

  12. 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.

  13. 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.

  14. 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)

  15. Fast cooling of trapped ions using the dynamical Stark shift gate

    E-Print Network [OSTI]

    A. Retzker; M. B. Plenio

    2006-07-27

    A laser cooling scheme for trapped ions is presented which is based on the fast dynamical Stark shift gate, described in [Jonathan etal, PRA 62, 042307]. Since this cooling method does not contain an off resonant carrier transition, low final temperatures are achieved even in traveling wave light field. The proposed method may operate in either pulsed or continuous mode and is also suitable for ion traps using microwave addressing in strong magnetic field gradients.

  16. Fluidic, Solid-State, and Hybrid Reconfiguration Techniques in a Frequency and Polarization Reconfigurable Antenna 

    E-Print Network [OSTI]

    Barrera, Joel Daniel

    2014-12-16

    Reconfigurable Antenna ISM Industrial, scientific, and medical I-V Current and voltage MEMS Microelectromechanical systems MS Microstrip PDMS Polydimethylsiloxane PIN P-type, intrinsic, N-type PN P-type and N-type PRA Polarization Reconfigurable Antenna...-8], and RF-MEMS [9-11] on or near the antenna surface. In [1], the authors achieve a widely frequency agile microstrip patch antenna by integrating three pairs of varactor diodes along with a broadband differential feeding scheme. Furthermore, the authors...

  17. 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.

  18. 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.

  19. 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.

  20. Combination of the single-double coupled cluster and the configuration interaction methods; application to barium, lutetium and their ions

    E-Print Network [OSTI]

    Dzuba, V A

    2014-01-01

    A version of the method of accurate calculations for few valence-electron atoms which combines linearized single-double coupled cluster method with the configuration interaction technique is presented. The use of the method is illustrated by calculations of the energy levels for Ba, Ba$^+$, Lu, Lu$^+$ and Lu$^{2+}$. Good agreement with experiment is demonstrated and comparison with previous version of the method (Safronova {\\em et al}, PRA {\\bf 80}, 012516 (2009)) is made.

  1. Exploration of High-Dimensional Scalar Function for Nuclear Reactor Safety Analysis and Visualization

    SciTech Connect (OSTI)

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

    2013-05-01

    The next generation of methodologies for nuclear reactor Probabilistic Risk Assessment (PRA) explicitly accounts for the time element in modeling the probabilistic system evolution and uses numerical simulation tools to account for possible dependencies between failure events. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. A challenge of dynamic PRA algorithms is the large amount of data they produce which may be difficult to visualize and analyze in order to extract useful information. We present a software tool that is designed to address these goals. We model a large-scale nuclear simulation dataset as a high-dimensional scalar function defined over a discrete sample of the domain. First, we provide structural analysis of such a function at multiple scales and provide insight into the relationship between the input parameters and the output. Second, we enable exploratory analysis for users, where we help the users to differentiate features from noise through multi-scale analysis on an interactive platform, based on domain knowledge and data characterization. Our analysis is performed by exploiting the topological and geometric properties of the domain, building statistical models based on its topological segmentations and providing interactive visual interfaces to facilitate such explorations. We provide a user’s guide to our software tool by highlighting its analysis and visualization capabilities, along with a use case involving dataset from a nuclear reactor safety simulation.

  2. A technique for human error analysis (ATHEANA)

    SciTech Connect (OSTI)

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

    1996-05-01

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

  3. 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.

  4. 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. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    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.

  5. 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.

  6. 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 environment such as RELAP-7. • Identify the risk-significant passive components, their failure modes and anticipated rates of degradation • Incorporate surveillance and maintenance activities and their effects into the plant state and into component aging progress. • Asses aging affects in a dynamic simulation environment 1. C. L. SMITH, V. N. SHAH, T. KAO, G. APOSTOLAKIS, “Incorporating Ageing Effects into Probabilistic Risk Assessment –A Feasibility Study Utilizing Reliability Physics Models,” NUREG/CR-5632, USNRC, (2001). 2. T. ALDEMIR, “A Survey of Dynamic Methodologies for Probabilistic Safety Assessment of Nuclear Power Plants, Annals of Nuclear Energy, 52, 113-124, (2013). 3. C. RABITI, A. ALFONSI, J. COGLIATI, D. MANDELLI and R. KINOSHITA “Reactor Analysis and Virtual Control Environment (RAVEN) FY12 Report,” INL/EXT-12-27351, (2012). 4. D. ANDERS et.al, "RELAP-7 Level 2 Milestone Report: Demonstration of a Steady State Single Phase PWR Simulation with RELAP-7," INL/EXT-12-25924, (2012).

  7. 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.

  8. Reverse osmosis desalination with osmotic polyelectrolyte intermediate 

    E-Print Network [OSTI]

    McConnell, Thomas Theodore

    1967-01-01

    by Loeb (27, 29) is the most promising membrane produced to date for reverse osmosis desalination. For production of potable water from saline water a salt rejection of 98. 6 per cent is necessary (15). In ac- tual pra-t. i. ce a greater salt... in comparison to a conventional reverse osmoti- cell with the same water flux. CHAP TER I I SURVEY OF THE LTTERATURE Research on desalination by reverse. osmotic means i. s a relatively new area of study. Most of the work in this field has been done...

  9. 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.

  10. Surveillance test interval optimization

    SciTech Connect (OSTI)

    Cepin, M.; Mavko, B. [Institut Jozef Stefan, Ljublijana (Slovenia)

    1995-12-31

    Technical specifications have been developed on the bases of deterministic analyses, engineering judgment, and expert opinion. This paper introduces our risk-based approach to surveillance test interval (STI) optimization. This approach consists of three main levels. The first level is the component level, which serves as a rough estimation of the optimal STI and can be calculated analytically by a differentiating equation for mean unavailability. The second and third levels give more representative results. They take into account the results of probabilistic risk assessment (PRA) calculated by a personal computer (PC) based code and are based on system unavailability at the system level and on core damage frequency at the plant level.

  11. 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.

  12. 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 using dynamic PRA methods. These efforts, documented in NUREG/CR-6901, NUREG/CR-6942, and NUREG/CR-6985, included a functional representation of the system's software but did not explicitly address failure modes caused by software defects or by inadequate design requirements. An important identified research need is to establish a commonly accepted basis for incorporating the behavior of software into digital I&C system reliability models for use in PRAs. To address this need, BNL is exploring the inclusion of software failures into the reliability models of digital I&C systems, such that their contribution to the risk of the associated NPP can be assessed.

  13. O Grotesco na Dramaturgia de Ariano Suassuna

    E-Print Network [OSTI]

    Telles, Narciso

    2002-04-01

    excluídos de nossa história. Arte Armoriai é o canto que vem do povo e que volta ao povo melhor do que veio."1 Como podemos notar, Ariano se baseia na ideia de circularidade cultural para sua elaboração estética onde o teatro torna-se porta voz do... partir do espaço onde a ação ocorre. A praça pede um vocabulário específico que nesta obra ganha mais vitalidade pelo uso de expressões grosseiras, libertadora do riso cómico que Suassuna maneja com extremo conhecimento. ANDREZA: Você vá pra merda...

  14. Review of the Diablo Canyon probabilistic risk assessment

    SciTech Connect (OSTI)

    Bozoki, G.E.; Fitzpatrick, R.G.; Bohn, M.P. [Sandia National Lab., Albuquerque, NM (United States); Sabek, M.G. [Atomic Energy Authority, Nuclear Regulatory and Safety Center, Cairo (Egypt); Ravindra, M.K.; Johnson, J.J. [EQE Engineering, San Francisco, CA (United States)

    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.

  15. All in the Family: Money, Kinship and Theravada Monasticism in Nepal

    E-Print Network [OSTI]

    Gellner, David N; LeVine, Sarah

    2007-01-01

    again lIsed her foreign contacts to secure a place for the girl in a Sri Lankan nunnery school which featured English ill its curriculum. Nuns sometimes make arrdngements for young male relatives and vice versa. A monk named Kos 11a tells how, as a young... . Bombay: Oxford University Press. Khare. R. S. ~984. The Untouchable as Himself: Ideology. Identity. and Pra?rna~lsm among Lucknow Cambarus. Cambridge: Cambridge University Press. March, Kathryn S. 1987. Hospitality. Women, and the Eflicacy of Beer. Food...

  16. 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.

  17. Results and insights of internal fire and internal flood analyses of the Surry Unit 1 Nuclear Power Plant during mid-loop operations

    SciTech Connect (OSTI)

    Chu, Tsong-Lun; Musicki, Z.; Kohut, P.

    1995-12-01

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). The objectives of the program are to assess the risks of severe accidents initiated during plant operational states (POSs) other than full power operation and to compare the estimated core damage frequencies (CDFs), 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 for internal events and a Level 1 PRA for seismically induced and internal fire and flood induced core damage sequences. This paper summarizes the results and highlights of the internal fire and flood analysis documented in Volumes 3 and 4 of NUREG/CR-6144 performed for the Surry plant during mid-loop operation.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. ISSUES ASSOCIATED WITH PROBABILISTIC FAILURE MODELING OF DIGITAL SYSTEMS

    SciTech Connect (OSTI)

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

    2004-09-19

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

  3. 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).

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. Component Fragility Research Program: Phase 1 component prioritization

    SciTech Connect (OSTI)

    Holman, G.S.; Chou, C.K.

    1987-06-01

    Current probabilistic risk assessment (PRA) methods for nuclear power plants utilize seismic ''fragilities'' - probabilities of failure conditioned on the severity of seismic input motion - that are based largely on limited test data and on engineering judgment. Under the NRC Component Fragility Research Program (CFRP), the Lawrence Livermore National Laboratory (LLNL) has developed and demonstrated procedures for using test data to derive probabilistic fragility descriptions for mechanical and electrical components. As part of its CFRP activities, LLNL systematically identified and categorized components influencing plant safety in order to identify ''candidate'' components for future NRC testing. Plant systems relevant to safety were first identified; within each system components were then ranked according to their importance to overall system function and their anticipated seismic capacity. Highest priority for future testing was assigned to those ''very important'' components having ''low'' seismic capacity. This report describes the LLNL prioritization effort, which also included application of ''high-level'' qualification data as an alternate means of developing probabilistic fragility descriptions for PRA applications.

  10. Impact of structural aging on seismic risk assessment of reinforced concrete structures in nuclear power plants

    SciTech Connect (OSTI)

    Ellingwood, B.; Song, J. [Johns Hopkins Univ., Baltimore, MD (United States). Dept. of Civil Engineering

    1996-03-01

    The Structural Aging Program is addressing the potential for degradation of concrete structural components and systems in nuclear power plants over time due to aging and aggressive environmental stressors. Structures are passive under normal operating conditions but play a key role in mitigating design-basis events, particularly those arising from external challenges such as earthquakes, extreme winds, fires and floods. Structures are plant-specific and unique, often are difficult to inspect, and are virtually impossible to replace. The importance of structural failures in accident mitigation is amplified because such failures may lead to common-cause failures of other components. Structural condition assessment and service life prediction must focus on a few critical components and systems within the plant. Components and systems that are dominant contributors to risk and that require particular attention can be identified through the mathematical formalism of a probabilistic risk assessment, or PRA. To illustrate, the role of structural degradation due to aging on plant risk is examined through the framework of a Level 1 seismic PRA of a nuclear power plant. Plausible mechanisms of structural degradation are found to increase the core damage probability by approximately a factor of two.

  11. 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.

  12. 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 requirements, including the need to operate in different coolant environments, higher operating temperatures, and longer operating cycles between planned refueling and maintenance outages. These features, along with the relative lack of operating experience for some of the proposed advanced designs, may limit the ability to estimate event probability and component POF with a high degree of certainty. Incorporating real-time estimates of component POF may compensate for a relative lack of established knowledge about the long-term component behavior and improve operational and maintenance planning and optimization. The particular eccentricities of advanced reactors and small modular reactors provide unique challenges and needs for advanced instrumentation, control, and human-machine interface (ICHMI) techniques such as enhanced risk monitors (ERM) in aSMRs. Several features of aSMR designs increase the need for accurate characterization of the real-time risk during operation and maintenance activities. A number of technical gaps in realizing ERM exist, and these gaps are largely independent of the specific reactor technology. As a result, the development of a framework for ERM would enable greater situational awareness regardless of the specific class of reactor technology. A set of research tasks are identified in a preliminary research plan to enable the development, testing, and demonstration of such a framework. Although some aspects of aSMRs, such as specific operational characteristics, will vary and are not now completely defined, the proposed framework is expected to be relevant regardless of such uncertainty. The development of an ERM framework will provide one of the key technical developments necessary to ensure the economic viability of aSMRs.

  13. 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 operation of systems and components important to safety as required in GDC 20. This paper provides an overview of the design process employed to develop a pre-conceptual FHR instrumentation architecture intended to lower plant capital and operational costs by minimizing reliance on expensive, safety related, safety-significant instrumentation through the use of inherent passive features of FHRs.

  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 a fire HRA. This process includes identification and definition of post-fire human failure events, qualitative analysis, quantification, recovery, dependency, and uncertainty. This document provides three approaches to quantification: screening, scoping, and detailed HRA. Screening is based on the guidance in NUREG/CR-6850, with some additional guidance for scenarios with long time windows. Scoping is a new approach to quantification developed specifically to support the iterative nature of fire PRA quantification. Scoping is intended to provide less conservative HEPs than screening, but requires fewer resources than a detailed HRA analysis. For detailed HRA quantification, guidance has been developed on how to apply existing methods to assess post-fire fire HEPs.

  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-specific estimates of POF into risk monitors, resulting in enhanced risk monitors that support optimized operation and maintenance of aSMRs.

  16. 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

  17. 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.

  18. Preliminary Risk Analysis of the LHC Cryogenic System (CERN-LHC-Project-Report-324)

    E-Print Network [OSTI]

    Chorowski, M; Riddone, G

    1999-01-01

    The Large Hadron Collider (LHC), presently under construction at CERN, will require a helium cryogenic system unprecedented in size and capacity, with more than 1600 superconducting magnets operating in superfluid helium and a total inventory of almost 100 tonnes of helium. The objective of the Preliminary Risk Analysis (PRA) is to identify all risks to personnel, equipment or environment resulting from failures that may accidentally occur within the cryogenic system of LHC in any phase of the machine operation, and that could not be eliminated by design. Assigning a gravity coefficient and one analyzing physical processes that will follow any of the recognised failure modes allows to single out worst case scenarios. Recommendations concerning lines of preventive and corrective defence, as well as for further detailed studies, are formulated.

  19. 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.

  20. Fifty years of progress in reactor safety

    SciTech Connect (OSTI)

    Okrent, D. (Univ. of California, Los Angeles (United States))

    1992-01-01

    This paper chronicles the major watershed occurrences in the evaluation of current reactor safety principles and concepts. The author covers such issues as the development of siting criteria in the early 1960s, development and design of engineered safety features and emergency cooling systems (ECCS), core meltdown scenarios, anticipated transients without scram (ATWS) issues, WASH-1400 Reactor Safety study employing probabilistic risk assessment (PRA) approaches, early PRAs conducted, development of safety goals in the 1980s, and reliability of AC power. Perhaps three of the most significant events related to operating reactors that occurred near the end of the 50-yr period were recognition of the problems of aging, completion of the NUREG-1150 study, and the development of a program on severe accident management.

  1. 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.

  2. 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)

  3. 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

  4. 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. (Science Applications International Corp., Albuquerque, NM (USA); Sandia National Labs., Albuquerque, NM (USA))

    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.

  5. 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.

  6. 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.

  7. Probabilistic risk assessment techniques help in identifying optimal equipment design for in-situ vitrification

    SciTech Connect (OSTI)

    Lucero, V.; Meale, B.M.; Purser, F.E.

    1990-01-01

    The analysis discussed in this paper was performed as part of the buried waste remediation efforts at the Idaho National Engineering Laboratory (INEL). The specific type of remediation discussed herein involves a thermal treatment process for converting contaminated soil and waste into a stable, chemically-inert form. Models of the proposed process were developed using probabilistic risk assessment (PRA) fault tree and event tree modeling techniques. The models were used to determine the appropriateness of the conceptual design by identifying potential hazards of system operations. Additional models were developed to represent the reliability aspects of the system components. By performing various sensitivities with the models, optimal design modifications are being identified to substantiate an integrated, cost-effective design representing minimal risk to the environment and/or public with maximum component reliability. 4 figs.

  8. Picture Invariance in Quantum Optics

    E-Print Network [OSTI]

    W. -Y. Hwang

    2008-08-12

    We clarify the controversy over the coherent-state (CS) versus the number-state (NS) pictures in quantum optics. The NS picture is equivalent to the CS picture, as long as the phases $\\phi$ in the laser fields are randomly distributed, as M{\\o}lmer argues [\\pra {\\bf 55}, 3195 (1997)]. However, the claim by Rudolph and Sanders [Phys. Rev. Lett. {\\bf 87}, 077903 (2001)] has a few gaps. First, they make an assumption that is not necessarily true in the calculation of a density operator involved with a two-mode squeezed state. We show that there exists entanglement in the density operator without defying the assumption that phases are randomly distributed. Moreover, using a concept of picture-invariance, we argue that it is not that criteria for quantum teleportation are not satisfied. We discuss an analogy between the controversy on the CS versus NS pictures to that on the heliocentric versus geocentric pictures.

  9. 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.

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. RAVEN. Dynamic Event Tree Approach Level III Milestone

    SciTech Connect (OSTI)

    Alfonsi, Andrea [Idaho National Lab. (INL), Idaho Falls, ID (United States); Rabiti, Cristian [Idaho National Lab. (INL), Idaho Falls, ID (United States); Mandelli, Diego [Idaho National Lab. (INL), Idaho Falls, ID (United States); Cogliati, Joshua [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinoshita, Robert [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    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.

  15. 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.

  16. 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.

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

    SciTech Connect (OSTI)

    Kunsman, David Marvin; Aldemir, Tunc (Ohio State University); Rutt, Benjamin (Ohio State University); Metzroth, Kyle (Ohio State University); Catalyurek, Umit (Ohio State University); Denning, Richard (Ohio State University); Hakobyan, Aram (Ohio State University); 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 such codes.) The results of this demonstration indicate that the approach can significantly reduce the resources required for Level 2 PRAs. From the phenomenological viewpoint, ADAPT can also treat the associated epistemic and aleatory uncertainties. This methodology can also be used for analyses of other complex systems. Any complex system can be analyzed using ADAPT if the workings of that system can be displayed as an event tree, there is a computer code that simulates how those events could progress, and that simulator code has switches to turn on and off system events, phenomena, etc. Using and applying ADAPT to particular problems is not human independent. While the human resources for the creation and analysis of the accident progression are significantly decreased, knowledgeable analysts are still necessary for a given project to apply ADAPT successfully. This research and development effort has met its original goals and then exceeded them.

  18. 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, 'Handbook of Human Reliability Analysis with Emphasis on Nuclear Power Plant Applications,' NUREG/CR-1278). This conversion produces the basic event risk of failure values required for the fault tree calculations. The fault tree is a deductive logic structure that corresponds to the operational nuclear MC&A system at a nuclear facility. The conventional Delphi process is a time-honored approach commonly used in the risk assessment field to extract numerical values for the failure rates of actions or activities when statistically significant data is absent.

  19. 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 possibility of corrosion cracking as in Davis-Besse and other plants. Latin America, as many other regions on the earth, needs water as much as electricity. IRIS has developed a water desalination co-generation design which can employ a variety of processes as dictated by local and economic conditions. Applications to the arid Brazilian Nord-Este and Mexican Nord-Oeste are being considered.

  20. MELCOR assessment at SNL

    SciTech Connect (OSTI)

    Kmetyk, L. N.

    1992-01-01

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants, being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission (USNRC). The entire spectrum of severe accident phenomena, including reactor coolant system and containment thermal/hydraulic response, core heatup, degradation and relocation, and fission product release and transport, is treated in MELCOR in a unified framework for both boiling water reactors (BWRS) and pressurized water reactors (PWRs). The MELCOR computer code has been developed to the point that it is now being successfully applied in severe accident analyses, particularly in probabilistic risk assessment (PRA) studies. MELCOR was the first of the severe accident analysis codes to undergo a formal peer review process. One of the major conclusions of the recent MELCOR Peer Review was the need for a more comprehensive and more systematic program of MELCOR assessment. A systematic program of code assessment provides a number of benefits, including: 1. guidance to the code developers in identification of areas where code improvements are needed (such as coding implementation errors in models, inappropriate or deficient models, missing models, excessive numerical sensitivities), 2. documented evidence to external observers, users, reviewers and project management that the code is modelling required phenomena correctly, and 3. increased general public acceptance that the code adequately treats issues related to public safety concerns.

  1. 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 Commission’s (NRC’s) 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 operator’s ability to diagnose the events and provide long term cooling.

  2. 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.)

  3. 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)

  4. Top-down and bottom-up definitions of human failure events in human reliability analysis

    SciTech Connect (OSTI)

    Boring, Ronald Laurids

    2014-10-01

    In the probabilistic risk assessments (PRAs) used in the nuclear industry, human failure events (HFEs) are determined as a subset of hardware failures, namely those hardware failures that could be triggered by human action or inaction. This approach is top-down, starting with hardware faults and deducing human contributions to those faults. Elsewhere, more traditionally human factors driven approaches would tend to look at opportunities for human errors first in a task analysis and then identify which of those errors is risk significant. The intersection of top-down and bottom-up approaches to defining HFEs has not been carefully studied. Ideally, both approaches should arrive at the same set of HFEs. This question is crucial, however, as human reliability analysis (HRA) methods are generalized to new domains like oil and gas. The HFEs used in nuclear PRAs tend to be top-down—defined as a subset of the PRA—whereas the HFEs used in petroleum quantitative risk assessments (QRAs) often tend to be bottom-up—derived from a task analysis conducted by human factors experts. The marriage of these approaches is necessary in order to ensure that HRA methods developed for top-down HFEs are also sufficient for bottom-up applications.

  5. 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.

  6. 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.

  7. Information Uncertainty to Compare Qualitative Reasoning Security Risk Assessment Results

    SciTech Connect (OSTI)

    Chavez, Gregory M [Los Alamos National Laboratory; Key, Brian P [Los Alamos National Laboratory; Zerkle, David K [Los Alamos National Laboratory; Shevitz, Daniel W [Los Alamos National Laboratory

    2009-01-01

    The security risk associated with malevolent acts such as those of terrorism are often void of the historical data required for a traditional PRA. Most information available to conduct security risk assessments for these malevolent acts is obtained from subject matter experts as subjective judgements. Qualitative reasoning approaches such as approximate reasoning and evidential reasoning are useful for modeling the predicted risk from information provided by subject matter experts. Absent from these approaches is a consistent means to compare the security risk assessment results. Associated with each predicted risk reasoning result is a quantifiable amount of information uncertainty which can be measured and used to compare the results. This paper explores using entropy measures to quantify the information uncertainty associated with conflict and non-specificity in the predicted reasoning results. The measured quantities of conflict and non-specificity can ultimately be used to compare qualitative reasoning results which are important in triage studies and ultimately resource allocation. Straight forward extensions of previous entropy measures are presented here to quantify the non-specificity and conflict associated with security risk assessment results obtained from qualitative reasoning models.

  8. Effect of tilting on turbulent convection: Cylindrical samples with aspect ratio $\\Gamma=0.50$

    E-Print Network [OSTI]

    Weiss, Stephan

    2015-01-01

    We report measurements of properties of turbulent thermal convection of a fluid with a Prandtl number $\\Pra=4.38$ in a cylindrical cell with an aspect ratio $\\Gamma=0.50$. The rotational symmetry was broken by a small tilt of the sample axis relative to gravity. Measurements of the heat transport (as expressed by the Nusselt number \\Nu), as well as of large-scale-circulation (LSC) properties by means of temperature measurements along the sidewall, are presented. In contradistinction to similar experiments using containers of aspect ratio $\\Gamma=1.00$ \\cite[]{ABN06} and $\\Gamma=0.50$ \\cite[]{CRCC04,SXX05,RGKS10}, we see a very small increase of the heat transport for tilt angles up to about 0.1 rad. Based on measurements of properties of the LSC we explain this increase by a stabilization of the single-roll state (SRS) of the LSC and a de-stabilization of the double-roll state (DRS) (it is known from previous work that the SRS has a slightly larger heat transport than the DRS). Further, we present quantitativ...

  9. 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.

  10. 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.

  11. Conductive and convective heat transfer in fluid flows between differentially heated and rotating cylinders

    E-Print Network [OSTI]

    Lopez, Jose M; Avila, Marc

    2015-01-01

    The flow of fluid confined between a heated rotating cylinder and a cooled stationary cylinder is a canonical experiment for the study of heat transfer in engineering. The theoretical treatment of this system is greatly simplified if the cylinders are assumed to be of infinite length or periodic in the axial direction, in which cases heat transfer occurs only through conduction as in a solid. We here investigate numerically heat transfer and the onset of turbulence in such flows by using both periodic and no-slip boundary conditions in the axial direction. We obtain a simple linear criterion that determines whether the infinite-cylinder assumption can be employed. The curvature of the cylinders enters this linear relationship through the slope and additive constant. For a given length-to-gap aspect ratio there is a critical Rayleigh number beyond which the laminar flow in the finite system is convective and so the behaviour is entirely different from the periodic case. The criterion does not depend on the Pra...

  12. 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.

  13. 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

  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 Commission’s (NRC’s) 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 component’s 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 NRC’s ability to include seismic hazards in risk assessments for operational events in support of the reactor oversight program (e.g., significance determination process).

  15. Proceedings of the 2004 international congress on advances in nuclear power plants - ICAPP'04

    SciTech Connect (OSTI)

    NONE

    2004-07-01

    The 2004 International Congress on Advances in Nuclear Power Plants (ICAPP'04) provides a forum for the industry to exchange the latest ideas and research findings on nuclear plants from all perspectives. This conference builds on the success of last year's meeting held in Cordoba, Spain, and on the 2002 inaugural meeting held in Hollywood, Florida. Because of the hard work of many volunteers from around the world, ICAPP'04 has been successful in achieving its goal. More than 325 invited and contributed papers/presentations are part of this ICAPP. There are 5 invited plenary sessions and 70 technical sessions with contributed papers. The ICAPP'04 Proceedings contain almost 275 papers prepared by authors from 25 countries covering topics related to advances in nuclear power plant technology. The program by technical track deals with: 1 - Water-Cooled Reactor Programs and Issues (Status of All New Water-Cooled Reactor Programs; Advanced PWRs: Developmental Stage I; Advanced PWRs: Developmental Stage II; Advanced PWRs: Basic Design Stage; Advanced BWRs; Economics, Regulation, Licensing, and Construction; AP1000); 2 - High Temperature Gas Cooled Reactors (Pebble Bed Modular Reactors; Very High Temperature Reactors; HTR Fuels and Materials; Innovative HTRs and Fuel Cycles); 3 - Long Term Reactor Programs and Strategies (Supercritical Pressure Water Reactors; Lead-Alloy Fast Reactors; Sodium and Gas Fast Reactors; Status of Advanced Reactor Programs; Non-classical Reactor Concepts); 4 - Operation, Performance, and Reliability Management (Information Technology Effect on Plant Operation; Operation, Maintenance and Reliability; Improving Performance and Reducing O and M Costs; Plant Modernization and Retrofits); 5 - Plant Safety Assessment and Regulatory Issues (LOCA and non-LOCA Analysis Methodologies; LOCA and non-LOCA Plant Analyses; In-Vessel Retention; Containment Performance and Hydrogen Control; Advances in Severe Accident Analysis; Advances in Severe Accident Management; Ex-Vessel Debris Coolability and Steam Explosion: Theory and Modeling; Ex-Vessel Debris Coolability and Steam Explosion: Experiments and Supporting Analysis; PRA and Risk-informed Decision Making: Methodology; PRA and Risk-informed Decision Making: Advances in Practice; Use of CFD in Plant Safety Assessment and Related Regulatory Issues; Development and Application of Severe Accident Analysis Code); 6 - Thermal Hydraulic Analysis and Testing (Advances in Two-Phase Flow and Heat Transfer; Advances in CHF and Rod Bundle Thermal Hydraulics; CFD Applications to Water, Liquid Metal, and Gas Reactors; Separate Effects Thermal Hydraulic Experiments and Analysis; Integral Systems Thermal Hydraulic Experiments; Benchmark Analysis and Assessment; Natural Circulation Thermal Hydraulics; Thermal Striping and Thermal Stratification Studies); 7 - Core and Fuel Cycle Concepts and Experiments (Innovations in Core Designs; Advances in Core Design Methodology and Experimental Benchmarking; Advanced Fuel Cycles, Recycling, and Actinide Transmutation; Out of Core Fuel Cycle Issues); 8 - Material and Structural Issues (Structural and Materials Modeling and Analysis; Testing and Analysis of Structures and Materials; Advanced Issues in Welding and Materials; Fuel Design and Irradiation Issues for Next Generation Plants; Materials' Issues for Next Generation Plants); 9 - Nuclear Energy and Sustainability Including Hydrogen, Desalination, and Other Applications (Nuclear Energy Sustainability and Desalination; Nuclear Energy Application - Hydrogen); 10 - Space Power and Propulsion (Space Nuclear Power and Propulsion Systems; Nuclear Thermal Propulsion Concepts; Test and Design Methods; Instrumentation for Space Nuclear Reactors; Materials for Space Reactor Concepts)

  16. Westinghouse Owners Group Risk-Informed Regulation Efforts: Options 2 and 3

    SciTech Connect (OSTI)

    Brown, Jason A.; Osterrieder, Robert A.; Lutz, Robert J. [Westinghouse Electric Company LLC (United States); Dingler, Maurice [Wolf Creek Nuclear Operating Company, Burlington, KS (United States); Ward, Lewis A. [Southern Nuclear Company (United States)

    2002-07-01

    The U.S. Nuclear Regulatory Commission (NRC) has initiated efforts to incorporate risk-informed methods to redefine the scope of the existing 10 CFR 50 regulations (Option 2) and to change the technical requirements of the regulations (Option 3). The overall objectives of these efforts are to enhance plant safety, provide a framework for risk-informed regulations, add flexibility to plant operations, and reduce regulatory burden. The Westinghouse Owners Group (WOG) has a variety of active programs in the risk-informed area, including a program in the Option 2 and Option 3 areas. These two programs will be summarized including the benefits and the technical approach. The purpose of Option 2 is to make changes to the overall scope of structures, systems and components (SSCs) covered by 10 CFR 50 requiring special treatment by formulating new risk-informed safety classification categories that are linked to current definitions of safety-related and important-to-safety. This initiative would permit possible changes to the current special treatment requirements based on risk insights. The Nuclear Energy Institute (NEI) has developed an Option 2 implementation guideline (NEI 00-04 Draft Revision B). The WOG has initiated a program to validate the NEI guideline and to provide an initial cost-benefit assessment of the revised categorization and treatment under Option 2 via trial application to two systems at both Surry Unit 1 and Wolf Creek. The WOG Option 2 program includes consideration of all of the components in the selected systems, regardless of whether or not they are modeled in the respective plant probabilistic risk assessment (PRA) studies. As a result, quantitative risk measures are not available for many of the components being considered. In this case, the WOG program will provide valuable input to the NEI guideline. Additionally, the WOG program extends the use of both of the dominant methodologies for risk-informed ISI (RI-ISI) to address repair and replacement activities of pressure-retaining items per Code Cases under development within ASME. Therefore, feedback is provided on the consideration of passive components for extending both the WOG and Electric Power Research Institute (EPRI) RI-ISI methodologies for piping to all pressure-retaining items. In the Option 3 area, the WOG Large Break Loss-of-Coolant Accident (LBLOCA) Redefinition program is a risk-informed approach to improve select regulations (10 CFR 50.46, Appendix A, and Appendix K) such that the plant licensing basis is focused on LOCA break sizes up to a new maximum size. The new maximum break size will replace the existing requirement to consider break sizes up to and including double ended breaks of the largest primary system piping. Plants will retain the capability to mitigate a break of the largest primary system piping as evaluated using realistic success criteria and assumptions. The WOG is also providing input to the development of a risk-informed 10 CFR 50.44 based on insights from the WOG Severe Accident Management Guidance and plant-specific PRA studies. (authors)

  17. COMBINED ACTIVE/PASSIVE DECAY HEAT REMOVAL APPROACH FOR THE 24 MWt GAS-COOLED FAST REACTOR

    SciTech Connect (OSTI)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01

    Decay heat removal at depressurized shutdown conditions has been regarded as one of the key areas where significant improvement in passive response was targeted for the GEN IV GFR over the GCFR designs of thirty years ago. It has been recognized that the poor heat transfer characteristics of gas coolant at lower pressures needed to be accommodated in the GEN IV design. The design envelope has therefore been extended to include a station blackout sequence simultaneous with a small break/leak. After an exploratory phase of scoping analysis in this project, together with CEA of France, it was decided that natural convection would be selected as the passive decay heat removal approach of preference. Furthermore, a double vessel/containment option, similar to the double vessel/guard vessel approach of the SFR, was selected as the means of design implementation to reduce the PRA risks of the depressurization accident. However additional calculations in conjunction with CEA showed that there was an economic penalty in terms of decay heat removal system heat exchanger size, elevation heights for thermal centers, and most of all in guard containment back pressure for complete reliance on natural convection only. The back pressure ranges complicated the design requirements for the guard containment. Recognizing that the definition of a loss-of-coolant-accident in the GFR is a misnomer, since gas coolant will always be present, and the availability of some driven blower would reduce fuel temperature transients significantly; it was decided instead to aim for a hybrid active/passive combination approach to the selected BDBA. Complete natural convection only would still be relied on for decay heat removal but only after the first twenty four hours after the initiation of the accident. During the first twenty four hour period an actively powered blower would be relied on to provide the emergency decay power removal. However the power requirements of the active blower/circulators would be kept low by maintaining a pressurized system coolant back pressure of {approx}7-8 bars through the design of the guard containment for such a design pressure. This approach is termed the medium pressure approach by both CEA and the US. Such a containment design pressure is in the range of the LWR experience, both PWRs and BWRs. Both metal containments and concrete guard containments are possible in this pressure range. This approach is then a time-at-risk approach as the power requirements should be low enough that battery/fuel cell banks without diesel generator start-up failure rate issues should be capable of providing the necessary power. Compressed gas sources are another possibility. A companion PRA study is being conducted to survey the reliability of such systems.

  18. High energy arcing fault fires in switchgear equipment : a literature review.

    SciTech Connect (OSTI)

    Nowlen, Steven Patrick; Brown, Jason W.; Wyant, Francis John

    2008-10-01

    In power generating plants, switchgear provide a means to isolate and de-energize specific electrical components and buses in order to clear downstream faults, perform routine maintenance, and replace necessary electrical equipment. These protective devices may be categorized by the insulating medium, such as air or oil, and are typically specified by voltage classes, i.e. low, medium, and high voltage. Given their high energy content, catastrophic failure of switchgear by means of a high energy arcing fault (HEAF) may occur. An incident such as this may lead to an explosion and fire within the switchgear, directly impact adjacent components, and possibly render dependent electrical equipment inoperable. Historically, HEAF events have been poorly documented and discussed in little detail. Recent incidents involving switchgear components at nuclear power plants, however, were scrupulously investigated. The phenomena itself is only understood on a very elementary level from preliminary experiments and theories; though many have argued that these early experiments were inaccurate due to primitive instrumentation or poorly justified methodologies and thus require re-evaluation. Within the past two decades, however, there has been a resurgence of research that analyzes previous work and modern technology. Developing a greater understanding of the HEAF phenomena, in particular the affects on switchgear equipment and other associated switching components, would allow power generating industries to minimize and possibly prevent future occurrences, thereby reducing costs associated with repair and downtime. This report presents the findings of a literature review focused on arc fault studies for electrical switching equipment. The specific objective of this review was to assess the availability of the types of information needed to support development of improved treatment methods in fire Probabilistic Risk Assessment (PRA) for nuclear power plant applications.

  19. Human Reliability Analysis for Small Modular Reactors

    SciTech Connect (OSTI)

    Ronald L. Boring; David I. Gertman

    2012-06-01

    Because no human reliability analysis (HRA) method was specifically developed for small modular reactors (SMRs), the application of any current HRA method to SMRs represents tradeoffs. A first- generation HRA method like THERP provides clearly defined activity types, but these activity types do not map to the human-system interface or concept of operations confronting SMR operators. A second- generation HRA method like ATHEANA is flexible enough to be used for SMR applications, but there is currently insufficient guidance for the analyst, requiring considerably more first-of-a-kind analyses and extensive SMR expertise in order to complete a quality HRA. Although no current HRA method is optimized to SMRs, it is possible to use existing HRA methods to identify errors, incorporate them as human failure events in the probabilistic risk assessment (PRA), and quantify them. In this paper, we provided preliminary guidance to assist the human reliability analyst and reviewer in understanding how to apply current HRA methods to the domain of SMRs. While it is possible to perform a satisfactory HRA using existing HRA methods, ultimately it is desirable to formally incorporate SMR considerations into the methods. This may require the development of new HRA methods. More practicably, existing methods need to be adapted to incorporate SMRs. Such adaptations may take the form of guidance on the complex mapping between conventional light water reactors and small modular reactors. While many behaviors and activities are shared between current plants and SMRs, the methods must adapt if they are to perform a valid and accurate analysis of plant personnel performance in SMRs.

  20. Potential for AP600 in-vessel retention through ex-vessel flooding

    SciTech Connect (OSTI)

    Rempe, J.L.; Knudson, D.L.; Allison, C.M.; Thinnes, G.L.; Atwood, C.L.

    1997-12-01

    External reactor vessel cooling (ERVC) is a new severe accident management strategy that involves flooding the reactor cavity to submerge the reactor vessel in an attempt to cool core debris that has relocated to the vessel lower head. Advanced and existing light water reactors (LWRs) are considering ERVC as an accident management strategy for in-vessel retention (IVR) of relocated debris. In the probabilistic risk assessment (PRA) for the AP600 design, Westinghouse credits ERVC for preventing vessel failure during postulated severe accidents with successful reactor coolant system (RCS) depressurization and reactor cavity flooding. To support the Westinghouse position on IVR, DOE contracted the University of California--Santa Barbara (UCSB) to produce the peer-reviewed report. To assist in the NRC`s evaluation of IVR of core melt by ex-vessel flooding of the AP6OO, the Idaho National Engineering and Environmental Laboratory (INEEL) was tasked to perform: An in-depth critical review of the UCSB study and the model that UCSB used to assess ERVC effectiveness; An in-depth review of the UCSB study peer review comments and of UCSB`s resolution method to identify areas where technical concerns weren`t addressed; and An independent analysis effort to investigate the impact of residual concerns on the margins to failure and conclusions presented in the UCSB study. This report summarizes results from these tasks. As discussed in Sections 1.1 and 1.2, INEEL`s review of the UCSB study and peer reviewer comments suggested that additional analysis was needed to assess: (1) the integral impact of peer reviewer-suggested changes to input assumptions and uncertainties and (2) the challenge present by other credible debris configurations. Section 1.3 summarized the corresponding analysis approach developed by INEEL. The remainder of this report provides more detailed descriptions of analysis methodology, input assumptions, and results.

  1. Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1. Volume 5: Analysis of core damage frequency from seismic events for plant operational state 5 during a refueling outage

    SciTech Connect (OSTI)

    Budnitz, R.J. [Future Resources Associates, Inc., Berkeley, CA (United States); Davis, P.R. [PRD Consulting (United States); Ravindra, M.K.; Tong, W.H. [EQE International, Inc., Irvine, CA (United States)

    1994-08-01

    In 1989 the US Nuclear Regulatory Commission (NRC) initiated an extensive program to examine carefully the potential risks during low-power and shutdown operations. The program included two parallel projects, one at Sandia National Laboratories studying a boiling water reactor (Grand Gulf), and the other at Brookhaven National Laboratory studying a pressurized water reactor (Surry Unit 1). Both the Sandia and Brookhaven projects have examined only accidents initiated by internal plant faults---so-called ``internal initiators.`` This project, which has explored the likelihood of seismic-initiated core damage accidents during refueling outage conditions, is complementary to the internal-initiator analyses at Brookhaven and Sandia. This report covers the seismic analysis at Grand Gulf. All of the many systems modeling assumptions, component non-seismic failure rates, and human effort rates that were used in the internal-initiator study at Grand Gulf have been adopted here, so that the results of the study can be as comparable as possible. Both the Sandia study and this study examine only one shutdown plant operating state (POS) at Grand Gulf, namely POS 5 representing cold shutdown during a refueling outage. This analysis has been limited to work analogous to a level-1 seismic PRA, in which estimates have been developed for the core-damage frequency from seismic events during POS 5. The results of the analysis are that the core-damage frequency for earthquake-initiated accidents during refueling outages in POS 5 is found to be quite low in absolute terms, less than 10{sup {minus}7}/year.

  2. Determining importance and grading of items and activities for the Yucca Mountain Project

    SciTech Connect (OSTI)

    DeKlever, R.; Verna, B.

    1993-12-31

    Raytheon Services Nevada (RSN), in support of the Department of Energy`s (DOE) Yucca Mountain Project, has been responsible for the Title 2 designs of the initial structures, systems, and components for the Exploratory Studies Facility (ESF), and the creation of the design output documents for the Surface-Based Testing (SBT) programs. The ESF and SBT programs are major scientific contributors to the overall site characterization program which will determine the suitability of Yucca Mountain to contain a proposed High Level Nuclear Waste (HLNW) repository. Accurate, traceable and objective characterization and testing documentation that is germane to the protection of public health and safety, and the environment, and that satisfies all the requirements of 10 CFR Part 60(1), must be established, evaluated and accepted. To assure that these requirements are satisfied, specific design functions and products, including items and activities depicted within respective design output documents, are subjected to the requirements of an NRC and DOE-approved Quality Assurance (QA) program. An evaluation (classification) is applied to these items and activities to determine their importance to radiological safety (ITS) and waste isolation (ITWI). Subsequently, QA program controls are selected (grading) for the items and activities. RSN has developed a DOE-approved classification process that is based on probabilistic risk assessment (PRA) techniques and that uses accident/impact scenarios. Results from respective performance assessment and test interference evaluations are also integrated into the classification analyses for various items. The methodology and results of the RSN classification and grading processes, presented herein, relative to ESF and SBT design products, demonstrates a solid, defensible methodological basis for classification and grading.

  3. Acute ethanol intake induces superoxide anion generation and mitogen-activated protein kinase phosphorylation in rat aorta: A role for angiotensin type 1 receptor

    SciTech Connect (OSTI)

    Yogi, Alvaro; Callera, Glaucia E.; Mecawi, André S.; Batalhão, Marcelo E.; Carnio, Evelin C.; Antunes-Rodrigues, José; Queiroz, Regina H.; Touyz, Rhian M.; Tirapelli, Carlos R.

    2012-11-01

    Ethanol intake is associated with increase in blood pressure, through unknown mechanisms. We hypothesized that acute ethanol intake enhances vascular oxidative stress and induces vascular dysfunction through renin–angiotensin system (RAS) activation. Ethanol (1 g/kg; p.o. gavage) effects were assessed within 30 min in male Wistar rats. The transient decrease in blood pressure induced by ethanol was not affected by the previous administration of losartan (10 mg/kg; p.o. gavage), a selective AT{sub 1} receptor antagonist. Acute ethanol intake increased plasma renin activity (PRA), angiotensin converting enzyme (ACE) activity, plasma angiotensin I (ANG I) and angiotensin II (ANG II) levels. Ethanol induced systemic and vascular oxidative stress, evidenced by increased plasma thiobarbituric acid-reacting substances (TBARS) levels, NAD(P)H oxidase?mediated vascular generation of superoxide anion and p47phox translocation (cytosol to membrane). These effects were prevented by losartan. Isolated aortas from ethanol-treated rats displayed increased p38MAPK and SAPK/JNK phosphorylation. Losartan inhibited ethanol-induced increase in the phosphorylation of these kinases. Ethanol intake decreased acetylcholine-induced relaxation and increased phenylephrine-induced contraction in endothelium-intact aortas. Ethanol significantly decreased plasma and aortic nitrate levels. These changes in vascular reactivity and in the end product of endogenous nitric oxide metabolism were not affected by losartan. Our study provides novel evidence that acute ethanol intake stimulates RAS activity and induces vascular oxidative stress and redox-signaling activation through AT{sub 1}-dependent mechanisms. These findings highlight the importance of RAS in acute ethanol-induced oxidative damage. -- Highlights: ? Acute ethanol intake stimulates RAS activity and vascular oxidative stress. ? RAS plays a role in acute ethanol-induced oxidative damage via AT{sub 1} receptor activation. ? Translocation of p47phox and MAPKs phosphorylation are downstream effectors. ? Acute ethanol consumption increases the risk for acute vascular injury.

  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. Supporting MELCOR calculations, Volume 6, Part 2

    SciTech Connect (OSTI)

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

    1995-03-01

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

  5. Quantification of margins and uncertainty for risk-informed decision analysis.

    SciTech Connect (OSTI)

    Alvin, Kenneth Fredrick

    2010-09-01

    QMU stands for 'Quantification of Margins and Uncertainties'. QMU is a basic framework for consistency in integrating simulation, data, and/or subject matter expertise to provide input into a risk-informed decision-making process. QMU is being applied to a wide range of NNSA stockpile issues, from performance to safety. The implementation of QMU varies with lab and application focus. The Advanced Simulation and Computing (ASC) Program develops validated computational simulation tools to be applied in the context of QMU. QMU provides input into a risk-informed decision making process. The completeness aspect of QMU can benefit from the structured methodology and discipline of quantitative risk assessment (QRA)/probabilistic risk assessment (PRA). In characterizing uncertainties it is important to pay attention to the distinction between those arising from incomplete knowledge ('epistemic' or systematic), and those arising from device-to-device variation ('aleatory' or random). The national security labs should investigate the utility of a probability of frequency (PoF) approach in presenting uncertainties in the stockpile. A QMU methodology is connected if the interactions between failure modes are included. The design labs should continue to focus attention on quantifying uncertainties that arise from epistemic uncertainties such as poorly-modeled phenomena, numerical errors, coding errors, and systematic uncertainties in experiment. The NNSA and design labs should ensure that the certification plan for any RRW is supported by strong, timely peer review and by an ongoing, transparent QMU-based documentation and analysis in order to permit a confidence level necessary for eventual certification.

  6. Offset, tilted dipole models of Uranian smooth high-frequency radio emission

    SciTech Connect (OSTI)

    Schweitzer, A.E.; Romig, J.H.; Evans, D.R.; Sawyer, C.B. (Radiophysics, Inc., Boulder, CO (USA)); Warwick, J.W. (Radiophysics, Inc., Boulder, CO (USA) Univ. of Colorado, Boulder (USA))

    1990-09-01

    During the Voyager 2 encounter with Uranus in January 1986, the Planetary Radio Astronomy (PRA) experiment detected a complex pattern of radio emissions. Two types of emissions were seen: smooth and bursty. The smooth emission has been divided into smooth high-frequency (SHF) and smooth low-frequency (SLF) components which are presumed to come from different sources because of their distinctly different characteristics. The SHF component is considered in this paper. The SHF emission has been modeled by many authors on OTD (offset, tilted dipole (Ness et al., 1986)) L shells ranging from 5 to 40. However, the bursts have been modeled at much higher L shells. The authors complete an OTD investigation of the SHF emission at high L shells within the range of the bursty source locations, and present a viable high L shell model. This model has fundamentally the same longitudinally symmetric net emission pattern in space as the L shell 5 model presented in Romig et al. (1987) and Barbosa (1988). However, they were unable to produce an acceptable model on intermediate L shells without restricting source longitude. They discuss the similarities and distinctions between their two models and the models of other authors. They believe that the high L shell model (and others similar to it) cannot account for the observed smoothness and periodicity of the SHF emissions because it has open field lines containing untrapped particles, which should produce more variable emission than that seen in the SHF data. Therefore, the authors prefer models at L shells less than 18, the boundary for closed field lines (Ness et al., 1986). They then discuss and contrast two models within this boundary: the L = 5 model and an L {approx} 12 model by Kaiser et al. (1987) and Farrell and Calvert (1989b). The main distinction between these two models is the longitudinal extent of the source location.

  7. Study of a Station Blackout Event in the PWR Plant

    SciTech Connect (OSTI)

    Ching-Hui Wu; Tsu-Jen Lin; Tsu-Mu Kao [Institute of Nuclear Energy Research P.O. Box 3-3, Longtan, 32500, Taiwan (China)

    2002-07-01

    On March 18, 2001, a PWR nuclear power plant located in the Southern Taiwan occurred a Station Blackout (SBO) event. Monsoon seawater mist caused the instability of offsite power grids. High salt-contained mist caused offsite power supply to the nuclear power plant very unstable, and forced the plant to be shutdown. Around 24 hours later, when both units in the plant were shutdown, several inadequate high cycles of bus transfer between 345 kV and 161 kV startup transformers degraded the emergency 4.16 kV switchgears. Then, in the Train-A switchgear room of Unit 1 occurred a fire explosion, when the degraded switchgear was hot shorted at the in-coming 345 kV breaker. Inadequate configuration arrangement of the offsite power supply to the emergency 4.16 kV switchgears led to loss of offsite power (LOOP) events to both units in the plant. Both emergency diesel generators (EDG) of Unit 1 could not be in service in time, but those of Unit 2 were running well. The SBO event of Unit 1 lasted for about two hours till the fifth EDG (DG-5) was lined-up to the Train-B switchgear. This study investigated the scenario of the SBO event and evaluated a risk profile for the SBO period. Guidelines in the SBO event, suggested by probabilistic risk assessment (PRA) procedures were also reviewed. Many related topics such as the re-configuration of offsite power supply, the addition of isolation breakers of the emergency 4.16 kV switchgears, the betterment of DG-5 lineup design, and enhancement of the reliability of offsite power supply to the PWR plant, etc., will be in further studies. (authors)

  8. Drilling and Production Testing the Methane Hydrate Resource Potential Associated with the Barrow Gas Fields

    SciTech Connect (OSTI)

    Steve McRae; Thomas Walsh; Michael Dunn; Michael Cook

    2010-02-22

    In November of 2008, the Department of Energy (DOE) and the North Slope Borough (NSB) committed funding to develop a drilling plan to test the presence of hydrates in the producing formation of at least one of the Barrow Gas Fields, and to develop a production surveillance plan to monitor the behavior of hydrates as dissociation occurs. This drilling and surveillance plan was supported by earlier studies in Phase 1 of the project, including hydrate stability zone modeling, material balance modeling, and full-field history-matched reservoir simulation, all of which support the presence of methane hydrate in association with the Barrow Gas Fields. This Phase 2 of the project, conducted over the past twelve months focused on selecting an optimal location for a hydrate test well; design of a logistics, drilling, completion and testing plan; and estimating costs for the activities. As originally proposed, the project was anticipated to benefit from industry activity in northwest Alaska, with opportunities to share equipment, personnel, services and mobilization and demobilization costs with one of the then-active exploration operators. The activity level dropped off, and this benefit evaporated, although plans for drilling of development wells in the BGF's matured, offering significant synergies and cost savings over a remote stand-alone drilling project. An optimal well location was chosen at the East Barrow No.18 well pad, and a vertical pilot/monitoring well and horizontal production test/surveillance well were engineered for drilling from this location. Both wells were designed with Distributed Temperature Survey (DTS) apparatus for monitoring of the hydrate-free gas interface. Once project scope was developed, a procurement process was implemented to engage the necessary service and equipment providers, and finalize project cost estimates. Based on cost proposals from vendors, total project estimated cost is $17.88 million dollars, inclusive of design work, permitting, barging, ice road/pad construction, drilling, completion, tie-in, long-term production testing and surveillance, data analysis and technology transfer. The PRA project team and North Slope have recommended moving forward to the execution phase of this project.

  9. Estimation of the Alpha Factor Parameters for the Emergency Diesel Generators of Ulchin Unit 3

    SciTech Connect (OSTI)

    Dae Il Kang; Sang Hoon Han

    2006-07-01

    Up to the present, the generic values of the Common cause failure (CCF) event parameters have been used in most PSA projects for the Korean NPPs. However, the CCF analysis should be performed with plant specific information to meet Category II of the ASME PRA Standard. Therefore, we estimated the Alpha factor parameters of the emergency diesel generator (EDG) for Ulchin Unit 3 by using the International Common-Cause Failure data Exchange (ICDE) database. The ICDE database provides the member countries with only the information needed for an estimation of the CCF parameters. The Ulchin Unit A3, pressurized water reactor, has two onsite EDGs and one alternate AC (AAC) diesel generator. The onsite EDGs of Unit 3 and 4 and the AAC are manufactured by the same company, but they are designed differently. The estimation procedure of the Alpha factor used in this study follows the approach of the NUREG/CR-5485. Since we did not find any qualitative difference between the target systems (two EDGs of Ulchin Unit 3) and the original systems (ICDE database), the applicability factor of each CCF event in the ICDE database was assumed to be 1. For the case of three EDGs including the AAC, five CCF events for the EDGs in the ICDE database were identified to be screened out. However, the detailed information for the independent events in the ICDE database is not presented. Thus, we assumed that the applicability factors for the CCF events to be screened out were, to be conservative, 0.5 and those of the other CCF events were 1. The study results show that the estimated Alpha parameters by using the ICDE database are lower than the generic values of the NUREG/CR-5497. The EDG system unavailability of the 1 out of 3 success criterion except for the supporting systems was calculated as 2.76 E-3. Compared with the system unavailability estimated by using the data of NUREG/CR-5497, it is decreased by 31.2%. (authors)

  10. System Effectiveness

    SciTech Connect (OSTI)

    Powell, Danny H; Elwood Jr, Robert H

    2011-01-01

    An effective risk assessment system is needed to address the threat posed by an active or passive insider who, acting alone or in collusion, could attempt diversion or theft of nuclear material. It is critical that a nuclear facility conduct a thorough self-assessment of the material protection, control, and accountability (MPC&A) system to evaluate system effectiveness. Self-assessment involves vulnerability analysis and performance testing of the MPC&A system. The process should lead to confirmation that mitigating features of the system effectively minimize the threat, or it could lead to the conclusion that system improvements or upgrades are necessary to achieve acceptable protection against the threat. Analysis of the MPC&A system is necessary to understand the limits and vulnerabilities of the system to internal threats. Self-assessment helps the facility be prepared to respond to internal threats and reduce the risk of theft or diversion of nuclear material. MSET is a self-assessment or inspection tool utilizing probabilistic risk assessment (PRA) methodology to calculate the system effectiveness of a nuclear facility's MPC&A system. MSET analyzes the effectiveness of an MPC&A system based on defined performance metrics for MPC&A functions based on U.S. and international best practices and regulations. A facility's MC&A system can be evaluated at a point in time and reevaluated after upgrades are implemented or after other system changes occur. The total system or specific subareas within the system can be evaluated. Areas of potential performance improvement or system upgrade can be assessed to determine where the most beneficial and cost-effective improvements should be made. Analyses of risk importance factors show that sustainability is essential for optimal performance. The analyses reveal where performance degradation has the greatest detrimental impact on total system risk and where performance improvements have the greatest reduction in system risk. The risk importance factors show the amount of risk reduction achievable with potential upgrades and the amount of risk reduction actually achieved after upgrades are completed. Applying the risk assessment tool gives support to budget prioritization by showing where budget support levels must be sustained for MC&A functions most important to risk. Results of the risk assessment are also useful in supporting funding justifications for system improvements that significantly reduce system risk.

  11. Fragility Analysis Methodology for Degraded Structures and Passive Components in Nuclear Power Plants - Illustrated using a Condensate Storage Tank

    SciTech Connect (OSTI)

    Nie, J.; Braverman, J.; Hofmayer, C.; Choun, Y.; Kim, M.; Choi, I.

    2010-06-30

    The Korea Atomic Energy Research Institute (KAERI) is conducting a five-year research project to develop a realistic seismic risk evaluation system which includes the consideration of aging of structures and components in nuclear power plants (NPPs). The KAERI research project includes three specific areas that are essential to seismic probabilistic risk assessment (PRA): (1) probabilistic seismic hazard analysis, (2) seismic fragility analysis including the effects of aging, and (3) a plant seismic risk analysis. Since 2007, Brookhaven National Laboratory (BNL) has entered into a collaboration agreement with KAERI to support its development of seismic capability evaluation technology for degraded structures and components. The collaborative research effort is intended to continue over a five year period. The goal of this collaboration endeavor is to assist KAERI to develop seismic fragility analysis methods that consider the potential effects of age-related degradation of structures, systems, and components (SSCs). The research results of this multi-year collaboration will be utilized as input to seismic PRAs. In the Year 1 scope of work, BNL collected and reviewed degradation occurrences in US NPPs and identified important aging characteristics needed for the seismic capability evaluations. This information is presented in the Annual Report for the Year 1 Task, identified as BNL Report-81741-2008 and also designated as KAERI/RR-2931/2008. The report presents results of the statistical and trending analysis of this data and compares the results to prior aging studies. In addition, the report provides a description of U.S. current regulatory requirements, regulatory guidance documents, generic communications, industry standards and guidance, and past research related to aging degradation of SSCs. In the Year 2 scope of work, BNL carried out a research effort to identify and assess degradation models for the long-term behavior of dominant materials that are determined to be risk significant to NPPs. Multiple models have been identified for concrete, carbon and low-alloy steel, and stainless steel. These models are documented in the Annual Report for the Year 2 Task, identified as BNL Report-82249-2009 and also designated as KAERI/TR-3757/2009. This report describes the research effort performed by BNL for the Year 3 scope of work. The objective is for BNL to develop the seismic fragility capacity for a condensate storage tank with various degradation scenarios. The conservative deterministic failure margin method has been utilized for the undegraded case and has been modified to accommodate the degraded cases. A total of five seismic fragility analysis cases have been described: (1) undegraded case, (2) degraded stainless tank shell, (3) degraded anchor bolts, (4) anchorage concrete cracking, and (5)a perfect combination of the three degradation scenarios. Insights from these fragility analyses are also presented.