Sample records for usive pra ctices

  1. Superdi usivity for a Brownian polymer in a continuous Gaussian environment

    E-Print Network [OSTI]

    Viens, Frederi G.

    environment (random medium) which can be brie y described as follows: the polymer itself, in the absenceSuperdi usivity for a Brownian polymer in a continuous Gaussian environment Sergio Bezerra Samy the asymptotic behavior of a one-dimen- sional Brownian polymer in random medium represented by a Gaussian eld W

  2. PRA and Risk Informed Analysis

    SciTech Connect (OSTI)

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

    2006-01-01T23:59:59.000Z

    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.

  3. To be submitted to Concrete Science and Engineering (August, 1999) New e ective medium theory for the di usivity or

    E-Print Network [OSTI]

    To be submitted to Concrete Science and Engineering (August, 1999) New e#11;ective medium theory for the di#11;usivity or conductivity of a multi-scale concrete microstructure model E.J. Garboczi National, CA 94551-9900 Abstract To attempt to represent concrete properly as a composite material, one must

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

    SciTech Connect (OSTI)

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

    1985-08-01T23:59:59.000Z

    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.

  5. Certification plan for safety and PRA codes

    SciTech Connect (OSTI)

    Toffer, H.; Crowe, R.D. (Westinghouse Hanford Co., Richland, WA (United States)); Ades, M.J. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1990-05-01T23:59:59.000Z

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

  6. Certification plan for safety and PRA codes

    SciTech Connect (OSTI)

    Toffer, H.; Crowe, R.D. [Westinghouse Hanford Co., Richland, WA (United States); Ades, M.J. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1990-05-01T23:59:59.000Z

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

  7. 01Aniversariantes do dia: 01 ALFREDO BUENO ASSISTENTE EM ADMINISTRACAO PRA/RU

    E-Print Network [OSTI]

    Paraná, Universidade Federal do

    SANTOS COZINHEIRO RA/AP 01 MARIA CLEONICE CAMPOS COZINHEIRO PRA/RU 01 MARIA ESMERALDA SANTOS DE MORAES

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

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

    P&RA CoP's Technical Exchange Meeting held on December 11-12, 2014 in Las Vegas, NV P&RA CoP's Technical Exchange Meeting held on December 11-12, 2014 in Las Vegas, NV P&RA CoP's...

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

    SciTech Connect (OSTI)

    Curtis Smith; Steven Prescott; Tony Koonce

    2014-04-01T23:59:59.000Z

    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.

  10. 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))

    1990-01-01T23:59:59.000Z

    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.

  11. Evolution of PRA methodology and insights since WASH-1400

    SciTech Connect (OSTI)

    Harper, F.T.; Camp, A.L.

    1986-01-01T23:59:59.000Z

    The US Nuclear Regulatory Commission (NRC) is preparing NUREG-1150 to examine the current perception of risk from a selected group of nuclear power plants. In support of NUREG-1150, Sandia National Laboratories (SNL) has directed the production of Level 1 Probabilistic Risk Assessments (PRAs) for the Surry, Sequoyah, Peach Bottom, and Grand Gulf nuclear power plants; additional studies are planned. The first four plants have been studied previously in either WASH-1400 or RSSMAP. The more recent studies suggest significant changes in perception of dominant accident sequences. In this paper the authors examine the changes in their perception of the likelihood of severe core damage accidents, in terms of both changes in PRA methodology and changes to the plants as a result of evolving regulations.

  12. Evolution of PRA methodology and insights since WASH-1400

    SciTech Connect (OSTI)

    Harper, F.T.; Camp, A.L.

    1986-01-01T23:59:59.000Z

    The Nuclear Regulatory Commission (NRC) is preparing NUREG-1150 to examine the current perception of risk from a selected group of nuclear power plants. In support of NUREG-1150, Sandia National Laboratories has directed the production of Level 1 Probabilistic Risk Assessments (PRAs) for the Surry, Sequoyah, Peach Bottom, and Grand Gulf nuclear power plants - additional studies are planned. The first four plants have been studied previously in either WASH-1400 or RSSMAP. The more recent studies suggest significant changes in our perception of dominant accident sequences. In this paper we will examine the changes in our perception of the likelihood of severe core damage accidents, in terms of both changes in PRA methodology and changes to the plants as a result of evolving regulations.

  13. Review of transient initiator frequencies of a BWR PRA

    SciTech Connect (OSTI)

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

    1985-01-01T23:59:59.000Z

    Estimation of transient initiator frequencies is important in the assessment of BWR core damage frequencies. The use of more up-to-date data bases and methodology as part of a peer review of a recent PRA, resulted in a different set of initiator frequencies. The impact on core damage frequencies is discussed. In addition, several related issues are addressed, such as the impact of excluding the first year of plant experience from the data base. It is concluded that the impact of the new up-to-date data base is significant, and more important than the effect of the use of the more rigorous two stage Bayesian method. The effect of ignoring the first year of experience has resulted in a reduction of merely 20% in the overall transient initiator frequency and 15% in the overall core damage frequency which is judged to be small.

  14. Failure record discounting in Bayesian analysis in Probabilistic Risk Assessment (PRA) : a space system application

    E-Print Network [OSTI]

    Lekkakos, Spyridon-Damianos

    2006-01-01T23:59:59.000Z

    In estimating a system-specific binomial probability of failure on demand in Probabilistic Risk Assessment (PRA), the corresponding number of observed failures may be not directly applicable due to design or procedure ...

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

    SciTech Connect (OSTI)

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

    1994-05-01T23:59:59.000Z

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

  16. Progesterone receptor isoforms PRA and PRB differentially contribute to breast cancer cell migration through interaction with focal adhesion kinase complexes

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 TITLE PAGE Progesterone receptor isoforms PRA and PRB differentially contribute to breast cancer A; PRB, progesterone receptor isoform B; ER, estrogen receptor ; P4, progesterone; R5020 (17 to impact mammary tumorigenesis, however the relative contribution of PRA and PRB isoforms in cancer cell

  17. Stabilization of Progesterone Receptor A and B isoforms by antiprogestin RU486 Identifies p38 and1 p42/44 MAPKs as Critical Regulators of PRA/PRB Ratio2

    E-Print Network [OSTI]

    Paris-Sud XI, Université de

    1 p42/44 MAPKs as Critical Regulators of PRA/PRB Ratio2 3 Abbreviated Title: Distinct MAPKs regulate PRA and PRB stability4 Precis: p38 and p42/44 MAPK stabilize PRA and PRB isoforms in a ligand: PR, progesterone receptor; PRA, progesterone receptor isoform A; PRB, progesterone27 receptor isoform

  18. Status of the Surry low power and shutdown PRA (probabilistic risk analysis)

    SciTech Connect (OSTI)

    Chu, T-L.; Luckas, W.; Musicki, Z.; Fitzpatrick, R.G.

    1990-01-01T23:59:59.000Z

    The Surry low power and shutdown probabilistic risk analysis (PRA) is an ongoing project at Brookhaven National Laboratory (BNL) to identify and quantify potential accident scenarios that may occur in a pressurized water reactor (PWR) during low power and shutdown. It was initiated as a result of various incidents and accidents that have occurred within the United States and overseas. The project involves review and evaluation of PWR experience at shutdown, identification of accident scenarios, determination of methods to mitigate the accidents, and performance a level 1 PRA. An evaluation of accident progression, source terms, and consequences has also been initiated. The results will be used to address issues related to shutdown conditions. The objective of this paper is to provide a progress report on the project, and to present the approach used as well as preliminary results of the ongoing and completed tasks. 14 refs., 1 fig., 5 tabs.

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

    SciTech Connect (OSTI)

    Sharp, D.A.

    1988-03-21T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Curtis Smith

    2013-09-01T23:59:59.000Z

    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.

  1. 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

    2011-03-01T23:59:59.000Z

    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.

  2. 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

    2011-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Not Available

    2004-09-01T23:59:59.000Z

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

  4. 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

    2013-06-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1988-01-01T23:59:59.000Z

    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.

  6. SUMMARY: PEST RISK ANALYSIS FOR PHYTOPHTHORA RAMORUM This summary presents the main features of a Pest Risk Analysis (PRA) which has been

    E-Print Network [OSTI]

    SUMMARY: PEST RISK ANALYSIS FOR PHYTOPHTHORA RAMORUM This summary presents the main features of a Pest Risk Analysis (PRA) which has been conducted on Phytophthora ramorum as the key deliverable from the EU-funded RAPRA Project. The PRA was prepared according to the EPPO Standard `Guidelines on Pest Risk

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

    SciTech Connect (OSTI)

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

    1985-01-01T23:59:59.000Z

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

  8. 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.

    1993-12-31T23:59:59.000Z

    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.

  9. When the Details Matter – Sensitivities in PRA Calculations That Could Affect Risk-Informed Decision-Making

    SciTech Connect (OSTI)

    Dana L. Kelly; Nathan O. Siu

    2010-06-01T23:59:59.000Z

    As the U.S. Nuclear Regulatory Commission (NRC) continues its efforts to increase its use of risk information in decision making, the detailed, quantitative results of probabilistic risk assessment (PRA) calculations are coming under increased scrutiny. Where once analysts and users were not overly concerned with figure of merit variations that were less than an order of magnitude, now factors of two or even less can spark heated debate regarding modeling approaches and assumptions. The philosophical and policy-related aspects of this situation are well-recognized by the PRA community. On the other hand, the technical implications for PRA methods and modeling have not been as widely discussed. This paper illustrates the potential numerical effects of choices as to the details of models and methods for parameter estimation with three examples: 1) the selection of the time period data for parameter estimation, and issues related to component boundary and failure mode definitions; 2) the selection of alternative diffuse prior distributions, including the constrained noninformative prior distribution, in Bayesian parameter estimation; and 3) the impact of uncertainty in calculations for recovery of offsite power.

  10. PRA In Design: Increasing Confidence in Pre-operational Assessments of Risks (Results of a Joint NASA/ NRC Workshop)

    SciTech Connect (OSTI)

    Robert Youngblood

    2010-06-01T23:59:59.000Z

    In late 2009, the National Aeronautics and Space Administration (NASA) and the U.S. Nuclear Regulatory Commission (NRC) jointly organized a workshop to discuss technical issues associated with application of risk assessments to early phases of system design. The workshop, which was coordinated by the Idaho National Laboratory, involved invited presentations from a number of PRA experts in the aerospace and nuclear fields and subsequent discussion to address the following questions: (a) What technical issues limit decision-makers’ confidence in PRA results, especially at a preoperational phase of the system life cycle? (b) What is being done to address these issues? (c) What more can be done? The workshop resulted in participant observations and suggestions on several technical issues, including the pursuit of non-traditional approaches to risk assessment and the verification and validation of risk models. The workshop participants also identified several important non-technical issues, including risk communication with decision makers, and the integration of PRA into the overall design process.

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

    SciTech Connect (OSTI)

    Barriere, M.T.; Luckas, W.J.; Brown, W.S. [Brookhaven National Lab., Upton, NY (United States); Cooper, S.E. [Science Applications International Corp., Reston, VA (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Bley, D.C. [PLG, Inc., Newport Beach, CA (United States)

    1993-12-31T23:59:59.000Z

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

  12. A fruitful experience at Frieda's, Inc. 

    E-Print Network [OSTI]

    Reeves, Melissa

    2000-01-01T23:59:59.000Z

    experience. CHAPTER ONE PRIMUSLABS. COM GROWER CERTIFICATION PROJECT Introduction Primuslabs. corn is a third party food safety certification organization based in Santa Maria, CA. They have operations in the United States and Mexico. Their services... student. 15 APPENDIX A Sample GAP Manual Developed with the Primuslabs. corn Document Development System LISSAS FARM LISSAS SANTA MARIA RANCH GOOD A GRICULTURAL PRA CTICES "GAP" AND STANDARD OPERA TING PROCEDURES ccSOP &s Prepared through www...

  13. A fruitful experience at Frieda's, Inc.

    E-Print Network [OSTI]

    Reeves, Melissa

    2000-01-01T23:59:59.000Z

    experience. CHAPTER ONE PRIMUSLABS. COM GROWER CERTIFICATION PROJECT Introduction Primuslabs. corn is a third party food safety certification organization based in Santa Maria, CA. They have operations in the United States and Mexico. Their services... student. 15 APPENDIX A Sample GAP Manual Developed with the Primuslabs. corn Document Development System LISSAS FARM LISSAS SANTA MARIA RANCH GOOD A GRICULTURAL PRA CTICES "GAP" AND STANDARD OPERA TING PROCEDURES ccSOP &s Prepared through www...

  14. 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

  15. Submitted to Geophysical and Astrophysical Fluid Dynamics Shear and Mixing in Oscillatory Doubly Di usive Convection

    E-Print Network [OSTI]

    Paparella, Francesco

    convection are found in the Earth's oceans, most notably, below the polar ice caps. There melting ice

  16. appraisal pra approach: Topics by E-print Network

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

    12; Is Java fit for large-scale computing? ffl Large and complex Kaltofen, Erich 5 Fractal, entropic and chaotic approaches to complex physiological time series analysis: a...

  17. 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

  18. ai techniques pra: Topics by E-print Network

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

    Last Page Topic Index 1 AI Programming Techniques Lab 3: Second order programming Computer Technologies and Information Sciences Websites Summary: G52APT AI Programming...

  19. California Public Records Act ("PRA"): In compliance with the PRA, the documents pertaining to agenda items, including attachments, which are presented to the City

    E-Print Network [OSTI]

    V) transmission entitlements (all located outside the City) pursuant to Vernon's Transmission Owner Tariff Adjustment for 2014 in accordance with Vernon's Transmission Owner Tariff and providing for tariff sheet attached revised Appendix I of Vernon's TO Tariff reflecting the TRBAA of positive $13,331; and d

  20. Microsoft Word - P&RA CoP Techncial Exchange Final Agenda 2014...

    Office of Environmental Management (EM)

    Annual Technical Exchange Meeting December 11 and 12, 2014 The National Atomic Testing Museum 755 E Flamingo Road, Las Vegas, NV 89119 Theme: Best Practices in Performance and Risk...

  1. Examples of the use of PRA in the design process and to support modifications

    SciTech Connect (OSTI)

    Johnson, D.H.; Bley, D.C.; Lin, J.C. [PLG, Inc., Newport Beach, CA (United States); Schueller, J.; van Otterloo, R.W. [Keuring van Elektrotechnische Materialen NV, Arnhem (Netherlands); Ramsey, C.T.; Linn, M.A. [Oak Ridge National Lab., TN (United States)

    1993-09-07T23:59:59.000Z

    Many, if not most, of the world`s commercial nuclear power plants have been the subject of plant-specific probabilistic safety assessments (PSAs). A growing number of other nuclear facilities as well as other types of industrial installations have been the focus of plant-specific PSAs. Such studies have provided valuable information concerning the nature of the fisk of the individual facility and have been utilized to identify opportunities to manage that risk. This paper explores the risk management activities associated with three diverse facilities to demonstrate the versatility of the use of PSA to support risk related decision making. The three facilities considered are a DOE research reactor with an extensive operating history, a proposed DOE research reactor in the advanced conceptual design phase and an offshore unmanned oil and gas installation.

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

    SciTech Connect (OSTI)

    Donald Helton; Hossein Esmaili; Robert Buell

    2011-03-01T23:59:59.000Z

    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.

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

    Office of Environmental Management (EM)

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

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

    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: Alternative1 First Use of Energy for All Purposes (Fuel and Nonfuel), 2002; Level: National5Sales for4,645 3,625 1,006 492 742 33 1112011 Strategic2 OPAM615_Cost Estimating Panel MicrosoftSeptember 15, SeptemberiOctoberMay 20,

  5. Sampling diffusive transition paths

    E-Print Network [OSTI]

    F. Miller III, Thomas

    2009-01-01T23:59:59.000Z

    Sampling di?usive transition paths Thomas F. Miller III ?the algorithm to sample the transition path ensemble for thedynamics I. INTRODUCTION Transition path sampling (TPS) is a

  6. 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...

  7. Probabilit`a e Statistica (LT in Matematica) Prof. P.Dai Pra, seconda prova parziale 12/03/2004.

    E-Print Network [OSTI]

    Vargiolu, Tiziano

    X(x)dx = + 0 etx FX (x)dx = lim b+ etx 1 - e- x2 2 b 0 - t b 0 etx 1 - e- x2 2 dx = lim b+ etb - etb e- b2 2

  8. DALLAS, TEXAS Localized Patterns in Homogeneous Networks of

    E-Print Network [OSTI]

    Moore, Peter K.

    ;usively Coupled Reactors Peter K. Moore, Werner Horsthemke SMU Math Report 04-002 DEPARTMENT; This research was partially supported by NSF Grant #DMS-0203154 y Department of Chemistry, SMU 1 #12; dynamics

  9. Updated 7/06/11 Section Numbers Course Type Instructional Method Site Code/Campus

    E-Print Network [OSTI]

    Karsai, Istvan

    Courses CON, HYB, IND, PRA, THS, etc. CLN CON TWY WEB WEB WEB RD1, RD2, RD3 IND,CON,PRA CON,IND,NCM CON, IND, HYB, THS, HYB WEB WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. NCM TWY

  10. Updated on 11/30/10 PHONE EMAIL

    E-Print Network [OSTI]

    Barros, Kipton PRA 7-6870 Behunin, Ryan PRA 5-2447 rbehunin Blume-Kohout, Robin DPF 6-0478 rbk Chien

  11. PRaVDA are delighted to announce that it has been selected as the winner in the Institute of Engineering and Technology (IET) Innovation Competition under the

    E-Print Network [OSTI]

    Wagner, Stephan

    of Engineering and Technology (IET) Innovation Competition under the model based engineering category. The Institution of Engineering and Technology's (IET) Innovation Awards recognise excellence across 16 categories Technology Innovation Award in 2012; and have grown to the point where they could float the company

  12. 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

  13. Estimation of effective diffusion coefficients in porous catalysts

    E-Print Network [OSTI]

    Kulkarni, Shrikant Ulhas

    1991-01-01T23:59:59.000Z

    'usivities were obtained for diR'usion of toluene in zeolites LaZSM-5, FeZSM-5 and BZSM-5. The corrected difl'usivities obtained for the zeolites showed a, dependence on the concentrat1on of' adsorbed species. Uptake experiments were conducted f' or studying... diffusion of n- hexane in a type II crystalline titanate, and the intracrystalline diffusivities were found to be independent of the adsorbate concentration. sv ACKNOWLEDGEMENT I would like to acknowledge my research advisor, Dr. R. G. Anthony...

  14. Revised 11/02/10 Section Numbers Course Type Instructional Method Site Code/Campus

    E-Print Network [OSTI]

    Karsai, Istvan

    WEB WEB WEB RD1, RD2, RD3 IND,CON,PRA CON,IND,NCM CON, IND, HYB, THS, HYB WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. WEB WEB TWY WEB CON, HYB, IND, PRA, THS, etc. NCM TWY WEB 23M Hospital Site or 23M 23M 23M 23M

  15. ccsd00003095, Quasi-geostrophic kinematic dynamos at low

    E-Print Network [OSTI]

    the geometry and the amplitude of the geomagnetic #12;eld [8]. All numerical models [9,10] have introduced;usivity) in the simulations. The current com- puter capacities limit the computation to magnetic Prandtl compared successfully their QG results with 3D calculations [16] and experimental measurement [17]. Low

  16. Massively Parallel Computation of Sti Propagating Combustion frontsMarc Garbey and Damien Tromeur-Dervout

    E-Print Network [OSTI]

    Garbey, Marc

    In this paper we study the computation of combustion fronts using MIMD archi- tecture. Our applications in gas models of combustion fronts: rst, a classical thermo-di usive model describing the combustion of a gasMassively Parallel Computation of Sti Propagating Combustion frontsMarc Garbey and Damien Tromeur

  17. J. Plasma Physics (2000), vol. 00, part 0, pp. 1{000 Copyright c

    E-Print Network [OSTI]

    Pohl, Martin Karl Wilhelm

    )). Progress in understanding the properties of the resulting energy spectra of the accel- erated particles 1 Turbulent adiabatic shock waves and di#11;usive particle acceleration By I. LE RCHEy, M. POHL of anomalous domains where the cosmic ray particle spectral index can be negative. All of these results

  18. fur Mathematik in den Naturwissenschaften

    E-Print Network [OSTI]

    to the internal variables fcg. The most distinctive feature of this operator is, that it acts on the densities f generated by di usive instabilities are by now quite well understood. In his pioneering work (cf. 24 wavelength; or in the case of dynamic patterns, their speed of propagation. The linearized analysis, however

  19. 1. Introduction This work is concerned with convection that is constrained by a vertical magnetic eld, moti-

    E-Print Network [OSTI]

    Rucklidge, Alastair

    in an in#12;nite layer. With periodic boundary conditions, travelling waves (TW) are also possible (Ruelle of the spot. Magneto- convection is oscillatory at onset if the magnetic di#11;usivity is small with sidewalls (Z 2 , or re ecting, boundary conditions), which allows only standing wave (SW) oscillations

  20. Mass conservative BDF-discontinuous Galerkin/explicit nite volume schemes for

    E-Print Network [OSTI]

    Recanati, Catherine

    Discontinuous Galerkin method. The kinematic wave equation governing the overland ow is discretized using porous medium, kinematic wave equation, Discontinuous Galerkin method, unstructured mesh PACS: 92.40.Kf-dimensional Richards' equation with a one-dimensional kinematic or di#11;usive wave approximation for the overland ow

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

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

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

  2. Training for Records and Information Management

    Broader source: Energy.gov [DOE]

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

  3. Farm records: a study of their relationship to farm management and federal income taxation.

    E-Print Network [OSTI]

    Warren, Steve Pearce

    1949-01-01T23:59:59.000Z

    charges incurred in acquirin' the 6o . ds vill m ice up the inventory value. In ce. . e of products produced on the f- rm th~ co. t b. si. will include co. t of raw m=. teri-l;, dir~ct labor =nd indirect expenses. l For ex mple, the co. t of . e..., to ~irect 1-)or for the entire farm ~nd pplying thi. ratio to the di- rect 1;, tor cost for the 10 acre plot, Indirect expense. m. , y be lioc. :t~d by any m thod which will conform to -cce. ted accountin~ pr ctice an& cle rly reflect. income, ' 1 Feder...

  4. A mathematical model of the productivity index of a well

    E-Print Network [OSTI]

    Khalmanova, Dinara Khabilovna

    2004-09-30T23:59:59.000Z

    Ibragimov for their continuing professional guidance and moral support in the last four years - this work would not have been possible without their help. All errors and inconsistencies remain my own. v NOMENCLATURE A - symmetric positive de nite matrix... of smooth coe cients CA - shape factor C - geometric characteristic of domain , de ned in terms of 0 h - thickness of the reservoir H1;2 - Sobolev space J - di usive capacity (productivity index) L - - elliptic operator, Lu = r Aru mesn - n...

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

    SciTech Connect (OSTI)

    Not Available

    1994-03-01T23:59:59.000Z

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

  6. Methodological Issues In Forestry Mitigation Projects: A Case Study Of Kolar District

    E-Print Network [OSTI]

    2008-01-01T23:59:59.000Z

    data Analysis Total cost Cost/tC Community ForestryFarm Forestry Activity Table 15: Transaction costprepared Community forestry Farm forestry Ecological PRA

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

    Energy Savers [EERE]

    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...

  8. Professional internship with American Cyanamid Company

    E-Print Network [OSTI]

    Haley, Perry W

    1993-01-01T23:59:59.000Z

    qt pr/A EPost A 8 v/v EPost A pt pr/A EPost A qt pr/A Epost pt pr/A EPost A 5 5 5 6 6 Pursuit Plus 28-0-0 NIS Pursuit Plus 28-0-0 Sunit II DG 0. 9 L 1. 0 L 1. 5 lh ai/A Epost A qt pr/A EPost A pt pr/A Epost A DG 0. 9 lh ai/A EPost A... 302 102 204 304 103 201 310 4 4 4 5 5 5 Pursuit Plus 28-0-0 NIS Counter pvS Pzowl 60DG 28-0-0 Sunit II Counter 3 EC 2. 5 L 1. 0 L 0. 25 15 G 8. 0 2 AS 4. 0 60 DG 1. 2 L 1. 0 L 1. 5 15 G 8. 0 pt pr/A lb pr/A qt pr/A pt pr...

  9. Memorandum of Understanding between the US Department of Energy...

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

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

  10. Performance Assessment Updates for Waste Isolation Pilot Plant...

    Office of Environmental Management (EM)

    December 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Performance Assessment Updates for Waste Isolation...

  11. Selected Failure Rate Values from ITER Safety Assessment

    Office of Scientific and Technical Information (OSTI)

    NFPA National Fire Protection Association NSSR Non-site Specific Safety Report OREDA Offshore Reliability Data Pa Pascal PHTS Primary Heat Transport System PRA Probabilistic...

  12. Comparison of Integrated Safety Analysis (ISA) and Probabilistic...

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

    Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 21711 Comparison of Integrated Safety Analysis (ISA) and...

  13. Comparison of Intergrated Safety Analysis (ISA) and Probabilistic...

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

    DC 20555-0001 SUBJECT: COMPARISON OF INTEGRATED SAFETY ANALYSIS (ISA) AND PROBABILISTIC RISK ASSESSMENT (PRA) FOR FUEL CYCLE FACILITIES Dear Chairman Jaczko: During the 580 th...

  14. National Marine Fisheries Service/NOAA, Commerce 222.101 (2) Result in improved operation of

    E-Print Network [OSTI]

    .S.C. 3501 et seq. (PRA). According to the PRA, a Federal agency may not con- duct or sponsor, and a person of the project works for electricity pro- duction. (c) When NMFS files with FERC the prescription that NMFS.102 Definitions. 222.103 Federal/state cooperation in the con- servation of endangered and threatened species

  15. 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 and an initial prioritisation of future investigations and possible risk reduction measures. The PRA comprised

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

    SciTech Connect (OSTI)

    Sasser, K.

    1994-06-01T23:59:59.000Z

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

  17. The social implications of mortuary remains at two Mimbres Mogollon sites in Grant County, New Mexico

    E-Print Network [OSTI]

    Spreen, Michele Rodes Kennedy

    1983-01-01T23:59:59.000Z

    are Attained and/or Le itimized by Means o Lineal Descent from the Dead i. e. Lineal Ties to Ancestors . Such Groups Will Maintain Formal Disposal Areas for the Exc usive Disposal o Their Dead, and Conversely Saxe 1970:119 author's italics. From... this hypothesis, Goldstein generates three subhypotheses. 1. To the degree that rights of corporate groups to use and/or control crucial but restricted resource(s) are attained and/or legitimized by lineal descent from the dead (i. e. , lineal ties...

  18. Augmenting Probabilistic Risk Assesment with Malevolent Initiators

    SciTech Connect (OSTI)

    Curtis Smith; David Schwieder

    2011-11-01T23:59:59.000Z

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

  19. A study of the breeding structure of the Aberdeen-Angus herd of the Agricultural and Mechanical College of Texas from 1906 to 1947

    E-Print Network [OSTI]

    Wheat, John David

    1951-01-01T23:59:59.000Z

    ~ this fsrti- cu';:r f?ni. ip " re, lace: -:. . . . ''1ii hi r. - u c: ~, . ' i. 'e rel:. &t ties? . ? tli. i. : . . u ei io ir Lh ~ fe . . 4iy, fr~re . ut?iJu tiie iiei;1 ii v. lhi, ct'uilg;, c' i; iio i, *" i ". . i?e iii ' cf tc1' if. tr- cd- eyetet...22o~s than;if sires taken at randu2a Lush (19~4) reported on a . horthorn hcril th. . t i;;;s clo. eu outside lclo c fox twenty ye-. rs. fhl: pr, -ctice brought about an Rvezag? xelaticnshlf of 4. per cent hex, , e u tlxe r ost outstanding Gine...

  20. Savannah River Site Probabilistic Risk Assessment high-level review

    SciTech Connect (OSTI)

    Not Available

    1990-02-01T23:59:59.000Z

    A review of the Savannah River Site (SRS) Probabilistic Risk Assessment (PRA) has been performed by a review committee organized by the US Department of Energy (DOE) and its contractor, EG&G Idaho, Inc. The High-Level Peer Review Committee (referred to as ``the Committee`` in this report) members are identified in Section 2. The main purpose of the review has been to provide assurance that the SRS PRA is responsive to safety issues associated with the restart and continued operation of the Savannah River reactors. The Committee members are all experienced practitioners of PRA, and several of the members have been deeply involved In a concurrent, detailed review of the SRS PRA. Source material and expertise available to the Committee included the SRS PRA document itself issued August 31. 1989, and Interaction with key PRA and plant experts at both the Savannah River Site and the Los Alamos National Laboratory (LANL), who had performed an independent PRA evaluation of the SRS K-reactor. The cooperation and support received from those connected with the review were outstanding.

  1. 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)

    2006-01-01T23:59:59.000Z

    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 ...

  2. Evaluation of the use of engineering judgements applied to analytical human reliablity analysis methods (HRA) 

    E-Print Network [OSTI]

    Kohlhepp, Katherine D.

    2006-04-12T23:59:59.000Z

    the use of engineering judgment applied to the quantification of post-initiator actions using the HRA Calculator. The Comanche Peak Steam Electric Station (CPSES) Level 1 Probabilistic Risk Assessment (PRA) HRA was used as a database for examples...

  3. Hormonal interactions in progesterone regulation of gonadotropin gene expression

    E-Print Network [OSTI]

    Ghochani, Yasmin

    2009-01-01T23:59:59.000Z

    receptor (PR): PRA and PRB, both encoded by the same gene,thought to be involved in transactivational function in PRB,making PRB a specific transcriptional regulator of target

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

    E-Print Network [OSTI]

    Debesse, Laurène

    2007-01-01T23:59:59.000Z

    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 ...

  5. Satellite System Safety Analysis Using STPA

    E-Print Network [OSTI]

    Dunn, Nicholas Connor

    2013-01-01T23:59:59.000Z

    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. ...

  6. Information Management and Supporting Documentation

    Broader source: Energy.gov [DOE]

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

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

    E-Print Network [OSTI]

    Huffman, Sarah Phillips

    1977-01-01T23:59:59.000Z

    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...

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

    E-Print Network [OSTI]

    Zhu, Yingli, 1976-

    2004-01-01T23:59:59.000Z

    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 ...

  9. Guidance for Incorporating Organizational Factors Into Nuclear Power Plant Risk Assessments - Phase 1 Workshop

    SciTech Connect (OSTI)

    J. Julius; A. Mosleh; M. Golay; V. Guthrie; J. Wreathall; A. Spurgin; B. Hannaman; D. Ziebell

    2002-12-31T23:59:59.000Z

    EPRI sponsored this study in order to help determine the influence of organizational factors on plant safety, risk, and economics. PRA tools provide excellent models for answering the question, ''How does change in an organizational factor impact the risk value?''

  10. 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)

    2010-01-01T23:59:59.000Z

    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 ...

  11. FIELD WORK (TD 609) Palsunda Village

    E-Print Network [OSTI]

    Sohoni, Milind

    ­ PRA, Household survey, Water Resources Survey, Road and Transport Survey, Energy Survey, Agriculture thank Prof. Puru Kulkarni, Prof. Milind Sohoni, Raj Desai Sir and Hemant for providing valuable inputs

  12. A New Interpretation of Flux Quantization Department of Physics, University of Puerto Rico

    E-Print Network [OSTI]

    Rico Humacao, PR 00791 Yong­Jihn Kim Department of Physics, University of Puerto Rico Mayaguez, PRA New Interpretation of Flux Quantization Mi­Ae Park Department of Physics, University of Puerto

  13. Nevada National Security Site Underground Test Area (UGTA) Flow...

    Office of Environmental Management (EM)

    December 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Nevada National Security Site Underground Test Area...

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

    Office of Environmental Management (EM)

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

  15. Nevada National Security Site Performance Assessment Updates...

    Office of Environmental Management (EM)

    December 11 and 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Nevada National Security Site Performance Assessment...

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

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

    NV December 11-12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Status Updates on the Performance and Risk Assessment...

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

    Office of Environmental Management (EM)

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

  18. Lessons Learned and Best Practices in Savannah River Site Saltstone...

    Office of Environmental Management (EM)

    Vegas, NV December 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Lessons Learned and Best Practices in Savannah River...

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

    Office of Environmental Management (EM)

    Estimation Since 2002 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation - Part 1 Video Presentation - Part 2 Risk Analysis and...

  20. Probabilistic Modeling and Phase 2 Decision Making at the West...

    Office of Environmental Management (EM)

    Meeting December 12, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Probabilistic Modeling and Phase 2 Decision Making at...

  1. Agenda - Interagency Steering Committee on Performance and Risk...

    Office of Environmental Management (EM)

    Phase 2 Studies, Dr. Zintars Zadins (Chenega) 4:30 - 6:00 pm Discussions, All Participants To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here....

  2. Androgen Receptor Repression of GnRH Gene Transcription

    E-Print Network [OSTI]

    Mellon, Pamela L.

    , and progesterone receptor (PR)A (4­6). The AR is a ligand-activated transcription factor, a member of the nuclear November 10, 2011 Abbreviations: AR, Androgen receptor; ChIP, chromatin immunoprecipitation; cs, char- coal

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

    E-Print Network [OSTI]

    Dulac, Nicolas, 1978-

    2007-01-01T23:59:59.000Z

    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 ...

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

    E-Print Network [OSTI]

    Ryan, Jason C. (Jason Christopher)

    2011-01-01T23:59:59.000Z

    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, ...

  5. 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 20-24, 2015 August 17-21, 2015 Fee: 275 Maximum enrollment: 28 30 hours AMA PRA Category 1 Credits(tm) This...

  6. Performance Assessment Community of Practice Technical Exchange

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

    Interagency Performance and Risk Assessment Community of Practice The Interagency Performance and Risk Assessment Community of Practice (P&RA CoP) was formed to provide a forum to...

  7. 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)

    2014-06-01T23:59:59.000Z

    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

  8. 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

    2013-11-01T23:59:59.000Z

    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.

  9. COMPONENT DEGRADATION SUSCEPTIBILITIES AS THE BASES FOR MODELING REACTOR AGING RISK

    SciTech Connect (OSTI)

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

    2010-07-18T23:59:59.000Z

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

  10. Validation of seismic probabilistic risk assessments of nuclear power plants

    SciTech Connect (OSTI)

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

    1994-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Uhrig, R.E.

    1987-01-01T23:59:59.000Z

    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).

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

    SciTech Connect (OSTI)

    Groth, Katrina M.; Swiler, Laura Painton

    2013-03-01T23:59:59.000Z

    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.

  13. Peer Review of NRC Standardized Plant Analysis Risk Models

    SciTech Connect (OSTI)

    Anthony Koonce; James Knudsen; Robert Buell

    2011-03-01T23:59:59.000Z

    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.

  14. SAPHIRE 8 Volume 3 - Users' Guide

    SciTech Connect (OSTI)

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

    2011-03-01T23:59:59.000Z

    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.

  15. Philosophy of ATHEANA

    SciTech Connect (OSTI)

    Bley, D.C.; Cooper, S.E.; Forester, J.A.; Kolaczkowski, A.M.; Ramey-Smith, A.; Thompson, C.M.; Whitehead, D.W.; Wreathall, J.

    1999-03-24T23:59:59.000Z

    ATHEANA, a second-generation Human Reliability Analysis (HRA) method integrates advances in psychology with engineering, human factors, and Probabilistic Risk Analysis (PRA) disciplines to provide an HRA quantification process and PRA modeling interface that can accommodate and represent human performance in real nuclear power plant events. The method uses the characteristics of serious accidents identified through retrospective analysis of serious operational events to set priorities in a search process for significant human failure events, unsafe acts, and error-forcing context (unfavorable plant conditions combined with negative performance-shaping factors). ATHEANA has been tested in a demonstration project at an operating pressurized water reactor.

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

    SciTech Connect (OSTI)

    Vierow, Karen; Aldemir, Tunc

    2009-09-10T23:59:59.000Z

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

  17. Fractal power spectra plotted upside-down Comment on ``Scaling of power spectrum of extinction events

    E-Print Network [OSTI]

    Kirchner, James W.

    Discussion Fractal power spectra plotted upside-down Comment on ``Scaling of power spectrum. Dimri and Pra- kash interpret their results as demonstrating a fractal pattern in the fossil record or not the underlying data are fractal. Similarly, their use of interpolated time series (in their ¢gures 1b,d, 2a,b, 3a

  18. From Maltreatment to Outcomes: Examining the Role of Explanations and Expectations as Mediators of the Maltreatment - Outcome Relation in Youth

    E-Print Network [OSTI]

    Makanui, Paul Kalani

    2008-01-01T23:59:59.000Z

    batery of measures. For the current study, caregivers completed the demographic form and the BASC ? I (PRA or PRC), while youth participants completed the CASQ ? R and the Y-LOT. Younger children (those under the age of 27 12) were read items from...

  19. (1) Likelihood mode of inference: definitions and results (i) Reference population and model.

    E-Print Network [OSTI]

    Frangakis, Constantine

    linear regression model with non additive effect. pr" ` "! CU# ` sr)¤ P u X }|e5 ~ ra 5 ¤ A B pr models. (1) Normal linear regression model with additive effect. pr"a`"! sU# ` sr)¤ P u X }| 5~ rf 5

  20. Quantum Information and Metrology with RF Traps at NIST D. J. Wineland, NIST, Boulder, CO

    E-Print Network [OSTI]

    Hensinger, Winfried

    . Knill (NIST)* D. Leibfried (NIST) D. Leibrandt (PostDoc, MIT) Y. Lin (student , CU) C. Ospelkaus (PD) # A. VanDevender (PD, U. Illinois) U. Warring (PD, Heidelberg) A. Wilson (guest researcher) D. J pseudopotential bumps (J. H. Wesenberg, PRA 78, 063410 (2008)) #12;Surface-electrode traps ~150 zone "racetrack

  1. Index of /~howardh

    E-Print Network [OSTI]

    PracticeExam1aSolutions.pdf, 09-Jan-2014 15:00, 2.0M. [ ], Section55parta.pdf, 05-Sep-2013 10:06, 544K. [ ], ex2_pra.pdf, 11-Nov-2013 15:18, 137K.

  2. A process for application of ATHEANA - a new HRA method

    SciTech Connect (OSTI)

    Parry, G.W. [NUS, Gaithersburg, MD (United States); Bley, D.C. [Buttonwood Consulting, Oakton, VA (United States); Cooper, S.E. [Science Applications International Corp., Reston, VA (United States)] [and others

    1996-10-01T23:59:59.000Z

    This paper describes the analytical process for the application of ATHEANA, a new approach to the performance of human reliability analysis as part of a PRA. This new method, unlike existing methods, is based upon an understanding of the reasons why people make errors, and was developed primarily to address the analysis of errors of commission.

  3. Electrical and Computer Engineering The George R. Brown School of Engineering

    E-Print Network [OSTI]

    Richards-Kortum, Rebecca

    Electrical and Computer Engineering The George R. Brown School of Engineering Chair Behnaam Aazhang in the PraCtiCe Scott Cutler Ray Simar, Jr. Gary Woods The Department of Electrical and Computer Engineering) and the bachelor of science in electrical engineering (BSEE). The BA degree #12;2 Departments / Electrical

  4. A Methodological Comparison of Monte Carlo Simulation and Epoch-Era Analysis for

    E-Print Network [OSTI]

    de Weck, Olivier L.

    techniques, morphological analysis, scenario planning · Semi-quantitative methods (can be used to initialize%) ­ Probabilistic risk assessment (PRA), Fault Tree Analysis (FTA), Hazards Analysis (HA), Failure modes and effectsA Methodological Comparison of Monte Carlo Simulation and Epoch-Era Analysis for Tradespace

  5. Approaches to uncertainty analysis in probabilistic risk assessment

    SciTech Connect (OSTI)

    Bohn, M.P.; Wheeler, T.A.; Parry, G.W.

    1988-01-01T23:59:59.000Z

    An integral part of any probabilistic risk assessment (PRA) is the performance of an uncertainty analysis to quantify the uncertainty in the point estimates of the risk measures considered. While a variety of classical methods of uncertainty analysis exist, application of these methods and developing new techniques consistent with existing PRA data bases and the need for expert (subjective) input has been an area of considerable interest since the pioneering Reactor Safety Study (WASH-1400) in 1975. This report presents the results of a critical review of existing methods for performing uncertainty analyses for PRAs, with special emphasis on identifying data base limitations on the various methods. Both classical and Baysian approaches have been examined. This work was funded by the US Nuclear Regulatory Commission in support of its ongoing full-scope PRA of the LaSalle nuclear power station. Thus in addition to the review, this report contains recommendations for a suitable uncertainty analysis methodology for the LaSalle PRA.

  6. Towards a 21st Century Postal Service John C. Panzar

    E-Print Network [OSTI]

    Bustamante, Fabián E.

    in volume · 20% off 2005 peak ­ Pension and Health Care overfunding · $75 BILLION (cumulative); $5 $7" ­ Pension overfunding ­ Health care overfunding · PAEA ­ Prevents real rate increases #12;Mail Volumes have ­ $12 billion per year by 1970. #12;US Postal Era II (19712006): The Postal Reform Act of 1970 (PRA

  7. The Unequal Burdens of Repatriation: A Gendered Analysis of the Transnational Migration of Mongolia's Kazakh Population

    E-Print Network [OSTI]

    Werner, Cynthia; Barcus, Holly R.

    esto tiene consecuencias financieras para las mujeres. Tercero, la migracio´n transnacional amplia las separaciones f?´sicas de los parientes natales que las mujeres ya sufren debido a las pra´cticas de parentesco Kazajo que enfatiza el linaje...

  8. An introduction Information and Records Management

    E-Print Network [OSTI]

    Hickman, Mark

    Bag 4800 Christchurch 8140 New Zealand www.canterbury.ac.nz Public RecordsAct ata glance · Only of the University. Public Records Act Information and Records Management at UC #12;Public RecordsAct2005 The Public Management team via records@canterbury.ac.nz. Archives New Zealand information on PRA, recordkeeping

  9. 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

  10. http://power.itp.ac.cn/~suncp/quantum.htm Fundamental physics for engineering quantum states

    E-Print Network [OSTI]

    Sun, Chang-Pu

    Zurek,Nature,412,712(2001); PRL, 89,170405 (2002) #12; Unitary Transformation ( ) ( ) (0) ( ) ( ) (0 Loschmidt Echo PRL 100, 100501 (2008) #12; Laflamme #12; , 20065130 Paz 2006 (Fidelity) 2007 At critical Point Probe of QPT by observing spectrum output of transmission Line #12; PRL06 PRL08, PRA09

  11. Dynamic reliability using entry-time approach for maintenance of nuclear power plants

    E-Print Network [OSTI]

    Wang, Shuwen

    2009-05-15T23:59:59.000Z

    V APPLICATION OF ENTRY-TIME MODEL TO NPP APPLICATIONS .......................................................................................116 5.1 Introduction to the Main Generator System...................................... 117 vii... of methods such as Probabilistic Risk Assessment (PRA), Markov models (cf. [1]), simulation [13] and other methods for risk assessment of dynamic systems [14] have been used to evaluate SSC (System, Structure and Component) reliability in NPP (Nuclear...

  12. Progesterone receptors: Form and function in brain Roberta Diaz Brinton a,b,*, Richard F. Thompson b,c,d

    E-Print Network [OSTI]

    Brinton, Roberta Diaz

    of Pharmacology and Pharmaceutical Sciences, University of Southern California, School of Pharmacy, 1985 Zonal (PR) that include the classic nuclear PRA and PRB receptors and splice variants of each, the seven of Pharmacology and Pharmaceutical Sciences, University of Southern California, School of Pharmacy, 1985 Zonal

  13. A Framework for Dynamic Safety and Risk Management Modeling in Complex Engineering Systems

    E-Print Network [OSTI]

    Leveson, Nancy

    analysis (FTA), and probabilistic risk analysis (PRA) rely on a chain-of-event paradigm of accident analysis or risk assessment techniques, such as failure modes and effect analysis (FMEA), fault tree and organizational boundaries. STAMP (System-Theoretic Accident Model and Processes) is a comprehensive accident

  14. Civil and Environmental Engineering The George R. Brown School of Engineering

    E-Print Network [OSTI]

    Richards-Kortum, Rebecca

    infrastructure and management (transportation systems, urban systems, soil mechanics, engineering economics129 Civil and Environmental Engineering The George R. Brown School of Engineering Chair Pedro. Wiesner adjunCt assistant Professor Karen Duston Professor in the PraCtiCe in Civil engineering management

  15. Civil and Environmental Engineering The George R. Brown School of Engineering

    E-Print Network [OSTI]

    Richards-Kortum, Rebecca

    infrastructure and management (transportation systems, urban systems, soil mechanics, engineering economics engineering management Joseph Cibor Ed Segner, III Professor in the PraCtiCe of environmental law James B137 Civil and Environmental Engineering The George R. Brown School of Engineering Chair Pedro

  16. SAPHIRE 8 Volume 7 - Data Loading

    SciTech Connect (OSTI)

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

    2011-03-01T23:59:59.000Z

    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.

  17. Pest Risk Analysis for Hymenoscyphus

    E-Print Network [OSTI]

    Pest Risk Analysis for Hymenoscyphus pseudoalbidus for the UK and the Republic of Ireland #12;2 PRA for Hymenoscyphus pseudoalbidus C.E. Sansford 23rd May 2013 Pest Risk Analysis Pest Risk Analysis for Hymenoscyphus (Kowalski and Holdenrieder, 2009). 1 Please cite this document as: Sansford, CE (2013). Pest Risk Analysis

  18. SAPHIRE 8 Volume 1 - Overview and Summary

    SciTech Connect (OSTI)

    C. L. Smith; S. T. Wood

    2011-03-01T23:59:59.000Z

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer (PC) running the Microsoft Windows operating system. SAPHIRE Version 8 is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). INL's primary role in this project is that of software developer and tester. However, INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users, who constitute a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events and quantify associated consequential outcome frequencies. Specifically, for nuclear power plant applications, SAPHIRE 8 can identify important contributors to core damage (Level 1 PRA) and containment failure during a severe accident which leads to releases (Level 2 PRA). It can be used for a PRA where the reactor is at full power, low power, or at shutdown conditions. Furthermore, it can be used to analyze both internal and external initiating events and has special features for managing models such as flooding and fire. It can also be used in a limited manner to quantify risk in terms of release consequences to the public and environment (Level 3 PRA). In SAPHIRE 8, the act of creating a model has been separated from the analysis of that model in order to improve the quality of both the model (e.g., by avoiding inadvertent changes) and the analysis. Consequently, in SAPHIRE 8, the analysis of models is performed by using what are called Workspaces. Currently, there are Workspaces for three types of analyses: (1) the NRC’s Accident Sequence Precursor program, where the workspace is called “Events and Condition Assessment (ECA);” (2) the NRC’s Significance Determination Process (SDP); and (3) the General Analysis (GA) workspace. Workspaces are independent of each other and modifications or calculations made within one workspace will not affect another. In addition, each workspace has a user interface and reports tailored for their intended uses. This report provides an overview of the functions and features available in SAPHIRE 8 and presents general instructions for using the software. Since SAPHIRE 8 expands upon Version 7, new and improved features will be discussed.

  19. Response of structures to energetic events for the Savannah River Site production reactors probabilistic risk assessment

    SciTech Connect (OSTI)

    Santa Cruz, S.M.; Smith, D.C. (Science Applications International Corp., Albuquerque, NM (United States)); Yau, W.F. (Westinghouse Savannah River Co., Aiken, SC (United States))

    1992-01-01T23:59:59.000Z

    The response of structures to energetic events postulated to arise in a probabilistic risk assessment (PRA) of a Savannah River Site (SRS) production reactor is addressed. Energetic events that arise in PRAs can damage structures and therefore have a significant influence on subsequent accident progression. Consequently, the structural response is important to the calculated risk of operating a plant. Difficulties are encountered, however, in the analysis of structural response of components to energetic loadings. First, the analysis of energetic events often does not provide well-defined static or dynamic loads acting on the structures. Secondly, risk assessments, by their nature, address a wide range of events that are not necessarily precisely defined. This paper describes an approach taken to develop the structural analysis required to support the PRA of the SRS production reactor, that overcomes these difficulties.

  20. Response of structures to energetic events for the Savannah River Site production reactors probabilistic risk assessment

    SciTech Connect (OSTI)

    Santa Cruz, S.M.; Smith, D.C. [Science Applications International Corp., Albuquerque, NM (United States); Yau, W.F. [Westinghouse Savannah River Co., Aiken, SC (United States)

    1992-10-01T23:59:59.000Z

    The response of structures to energetic events postulated to arise in a probabilistic risk assessment (PRA) of a Savannah River Site (SRS) production reactor is addressed. Energetic events that arise in PRAs can damage structures and therefore have a significant influence on subsequent accident progression. Consequently, the structural response is important to the calculated risk of operating a plant. Difficulties are encountered, however, in the analysis of structural response of components to energetic loadings. First, the analysis of energetic events often does not provide well-defined static or dynamic loads acting on the structures. Secondly, risk assessments, by their nature, address a wide range of events that are not necessarily precisely defined. This paper describes an approach taken to develop the structural analysis required to support the PRA of the SRS production reactor, that overcomes these difficulties.

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

    SciTech Connect (OSTI)

    Suzuki, Mitsutoshi; Mochiji, Toshiro [Department of Science and Technology for Nuclear Material Management, Japan Atomic Energy Agency, 2-4 Shirane, Shirakata, Tokai-mura, Naka-gun, Ibaraki 319-1193 (Japan)

    2014-09-30T23:59:59.000Z

    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.

  2. SAPHIRE 8 Software Quality Assurance Plan

    SciTech Connect (OSTI)

    Curtis Smith

    2010-02-01T23:59:59.000Z

    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.

  3. 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 [Entergy Nuclear South (United States)

    2006-07-01T23:59:59.000Z

    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)

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

    SciTech Connect (OSTI)

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

    2012-09-20T23:59:59.000Z

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

  5. University of Maryland Condensed Matter Theory Center

    E-Print Network [OSTI]

    Scarola, Vito

    mechanically screen Coulomb interaction Non-interacting ·Exact in two limits: } } Jain PRL `89 Yi,Fertig PRB Oscillates ·High Overlap Harju et al. PRL `02 Burkard et al. PRB `99 Hu,Das Sarma PRA `00 #12;Empirical Two-like"Noise) S1 S2 N N Nuclear flip-flop Manipulation/ Detection Khaetskii,Nazarov PRB '01 DeSousa,Das Sarma PRB

  6. EECS 310, Fall 2011 Instructor: Nicole Immorlica

    E-Print Network [OSTI]

    Immorlica, Nicole

    B be an event such that Pr{B} > 0. Define a function PrB{.} on outcome w S by the rule: PrB{w} = {Pr{w}/ Pr B if w B; 0 if w B}. (a) (10 points) Prove that PrB{.} is also a probability functio on S according to Defi- nition 14.4.2 in the MIT notes. #12;(b) (10 points) Prove that Pr B {A} = Pr{A B} Pr{B

  7. Dynamic Event Tree Analysis Through RAVEN

    SciTech Connect (OSTI)

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

    2013-09-01T23:59:59.000Z

    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.

  8. Cortisol-induced inhibition of ovine renin and aldosterone responses to hypotension

    SciTech Connect (OSTI)

    Wood, C.E.; Silbiger, J.

    1987-03-01T23:59:59.000Z

    Previous studies from this laboratory have demonstrated that in preterm fetal sheep increases in plasma cortisol (F) concentration equal in amplitude to fetal F stress responses suppress plasma renin activity (PRA). The purpose of this study was to investigate the possibility that this negative interaction exists in adult sheep. Cortisol was measured by radioimmunoassay. Five conscious ewes with chronically prepared carotid arterial loops were infused intravenously with F or vehicle for 5 h. One hour after the end of F or vehicle infusion, renin secretion was stimulated by hypotension produced by infusion of sodium nitroprusside. F infusion increased plasma F; during vehicle infusion plasma F did not change. F infusion decreased hematocrit from 29 +/- 2 to 26 +/- 1%. Basal PRA in vehicle- and F-infused groups were 0.4 +/- 0 and 0.2 +/- 0.1 ng angiotensin I-ml/sup -1/-h/sup -1/ and did not change. In vehicle-infused ewes, PRA increased from 0.4 +/- 0 to 4.6 +/- 0.4 and plasma aldosterone from 26.0 +/- 1.0 to 173.1 +/- 21.8 pg/ml, while in F-infused ewes, PRA increased from 0.2 +/- 1 to 3.3 +/- 0.4 ng angiotensin I-ml/sup -1/-h/sup -1/ and aldosterone from 25.0 +/- 0 to 48.2 +/- 23.2 pg/ml, significantly smaller responses. These results suggest that repeated stress may modulate the responses of the renin-angiotensin system in this species.

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

    SciTech Connect (OSTI)

    Zhegang Ma

    2013-09-01T23:59:59.000Z

    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.

  10. Exploration of high-dimensional scalar function for nuclear reactor safety analysis and visualization

    SciTech Connect (OSTI)

    Maljovec, D.; Wang, B.; Pascucci, V. [Scientific Computing and Imaging Institute, University of Utah (United States); Bremer, P. T. [Lawrence Livermore National Laboratory (United States); Pernice, M.; Mandelli, D.; Nourgaliev, R. [Idaho National Laboratory (United States)

    2013-07-01T23:59:59.000Z

    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 data from a nuclear reactor safety simulation. (authors)

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

    SciTech Connect (OSTI)

    Lloyd, R C; Moffitt, N E; Gore, B F; Vo, T V; Vehec, T A [Pacific Northwest Lab., Richland, WA (United States)

    1993-02-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Pugh, R.; Gore, B.F.; Vo, T.V.; Moffitt, N.E. (Pacific Northwest Lab., Richland, WA (United States))

    1991-09-01T23:59:59.000Z

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

  13. NRC support for the Kalinin (VVER) probabilistic risk assessment

    SciTech Connect (OSTI)

    Bley, D. [Buttonwood Consulting, Inc. (United States); Diamond, D.J.; Chu, T.L.; Azarm, A.; Pratt, W.T. [Brookhaven National Lab., Upton, NY (United States); Johnson, D. [PLG, Inc. (United States); Szukiewicz, A.; Drouin, M.; El-Bassioni, A.; Su, T.M. [Nuclear Regulatory Commission (United States)

    1998-12-31T23:59:59.000Z

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

  14. A technique for human error analysis (ATHEANA)

    SciTech Connect (OSTI)

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

    1996-05-01T23:59:59.000Z

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

  15. NRC SUPPORT FOR THE KALININ (VVER) PROBABILISTIC RISK ASSESSMENT

    SciTech Connect (OSTI)

    BLEY,D.; DIAMOND,D.J.; CHU,T.L.; AZARM,A.; PRATT,W.T.; JOHNSON,D.; SZUKIEWICZ,A.; DROUIN,M.; EL-BASSIONI,A.; SU,T.M.

    1998-10-26T23:59:59.000Z

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

  16. 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

    2013-05-01T23:59:59.000Z

    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.

  17. A process for risk-focused maintenance

    SciTech Connect (OSTI)

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

    1991-03-01T23:59:59.000Z

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

  18. Advanced Test Reactor probabilistic risk assessment methodology and results summary

    SciTech Connect (OSTI)

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

    1992-01-01T23:59:59.000Z

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

  19. Assessing conformance to safety goals using nonparametric empirical Bayes methods: A nuclear reactor application

    SciTech Connect (OSTI)

    Martz, H.F.; Johnson, J.W. [Los Alamos National Lab., NM (United States)

    1997-01-01T23:59:59.000Z

    Nonparametric empirical Bayes methods are used to develop decision criteria for use in deciding whether the risk of a given facility is compatible with a corresponding specified quantitative safety goal. The criteria utilize the uncertain results of a probabilistic risk assessment (PRA) and are derived from an empirical Bayes point of view in which the results from a set of similar facilities are used to estimate the population variability curve (PVC) for the parameter of interest. The PVC is estimated nonparametrically in the sense that the distributional family to which the PVC belongs is completely unknown and unspecified. For the assumed model, the method guarantees that all facilities ultimately accepted as being compatible with the goal have a prespecified exact assurance probability that the goal is not exceeded. The method also accounts for two possible biases in the PRA results. Criteria are developed for use in assessing the compatibility of nuclear power plant PRA-produced severe core damage frequency estimates with a corresponding subsidiary objective.

  20. Developing and evaluating distributions for probabilistic human exposure assessments

    SciTech Connect (OSTI)

    Maddalena, Randy L.; McKone, Thomas E.

    2002-08-01T23:59:59.000Z

    This report describes research carried out at the Lawrence Berkeley National Laboratory (LBNL) to assist the U. S. Environmental Protection Agency (EPA) in developing a consistent yet flexible approach for evaluating the inputs to probabilistic risk assessments. The U.S. EPA Office of Emergency and Remedial Response (OERR) recently released Volume 3 Part A of Risk Assessment Guidance for Superfund (RAGS), as an update to the existing two-volume set of RAGS. The update provides policy and technical guidance on performing probabilistic risk assessment (PRA). Consequently, EPA risk managers and decision-makers need to review and evaluate the adequacy of PRAs for supporting regulatory decisions. A critical part of evaluating a PRA is the problem of evaluating or judging the adequacy of input distributions PRA. Although the overarching theme of this report is the need to improve the ease and consistency of the regulatory review process, the specific objectives are presented in two parts. The objective of Part 1 is to develop a consistent yet flexible process for evaluating distributions in a PRA by identifying the critical attributes of an exposure factor distribution and discussing how these attributes relate to the task-specific adequacy of the input. This objective is carried out with emphasis on the perspective of a risk manager or decision-maker. The proposed evaluation procedure provides consistency to the review process without a loss of flexibility. As a result, the approach described in Part 1 provides an opportunity to apply a single review framework for all EPA regions and yet provide the regional risk manager with the flexibility to deal with site- and case-specific issues in the PRA process. However, as the number of inputs to a PRA increases, so does the complexity of the process for calculating, communicating and managing risk. As a result, there is increasing effort required of both the risk professionals performing the analysis and the risk manager reviewing it. For deterministic risk assessments, the use of default inputs has improved the ease and the consistency of both performing and reviewing assessments. By analogy, it is expected that similar advantage will be seen in the field of probabilistic risk assessment through the introduction of default distributions. In Part 2 of this report, we consider when a default distribution might be appropriate for use in PRA and work towards development of recommended task-specific distributions for several frequently used exposure factors. An approach that we develop using body weight and exposure duration as case studies offers a transparent way for developing task-specific exposure factor distributions. A third case study using water intake highlights the need for further study aimed at improving the relevance of ''short-term'' data before recommendations on task-specific distributions of water intake can be made.

  1. 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

    2014-11-01T23:59:59.000Z

    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).

  2. Solution NMR and X-ray Crystal Structures of Membrane-associated Lipoprotein-17 Domain Reveal a Novel Fold

    SciTech Connect (OSTI)

    R Mani; S Vorobiev; G Swapna; H Neely; H Janjua; C Ciccosanti; D Xiao; J Hunt; G Montelione; et al.

    2011-12-31T23:59:59.000Z

    The conserved Lipoprotein-17 domain of membrane-associated protein Q9PRA0{_}UREPA from Ureaplasma parvum was selected for structure determination by the Northeast Structural Genomics Consortium, as part of the Protein Structure Initiative's program on structure-function analysis of protein domains from large domain sequence families lacking structural representatives. The 100-residue Lipoprotein-17 domain is a 'domain of unknown function' (DUF) that is a member of Pfam protein family PF04200, a large domain family for which no members have characterized biochemical functions. The three-dimensional structure of the Lipoprotein-17 domain of protein Q9PRA0{_}UREPA was determined by both solution NMR and by X-ray crystallography at 2.5 {angstrom}. The two structures are in good agreement with each other. The domain structure features three {alpha}-helices, {alpha}1 through {alpha}3, and five {beta}-strands. Strands {beta}1/{beta}2, {beta}3/{beta}4, {beta}4/{beta}5 are anti-parallel to each other. Strands {beta}1 and {beta}2 are orthogonal to strands {beta}3, {beta}4, {beta}5, while helix {alpha}3 is formed between the strands {beta}3 and {beta}4. One-turn helix {alpha}2 is formed between the strands {beta}1 and {beta}2, while helix {alpha}1 occurs in the N-terminal polypeptide segment. Searches of the Protein Data Bank do not identify any other protein with significant structural similarity to Lipoprotein-17 domain of Q9PRA0{_}UREPA, indicating that it is a novel protein fold.

  3. Risk-Informed Safety Margin Characterization Methods Development Work

    SciTech Connect (OSTI)

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

    2014-09-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1995-03-01T23:59:59.000Z

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

  5. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS.

    SciTech Connect (OSTI)

    BRAVERMAN,J.I.; MORANTE,R.J.; XU,J.; HOFMAYER,C.H.; SHAUKAT,S.K.

    2003-03-17T23:59:59.000Z

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister (MPC) that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel.

  6. IMPACT ANALYSIS OF SPENT FUEL DRY CASKS UNDER ACCIDENTAL DROP SCENARIOS.

    SciTech Connect (OSTI)

    BRAVERMAN,J.I.; MORANTE,R.J.; XU,J.; HOFMAYER,C.H.; SHAUKAT,S.K.

    2003-08-17T23:59:59.000Z

    A series of analyses were performed to assess the structural response of spent nuclear fuel dry casks subjected to various handling and on-site transfer events. The results of these analyses are being used by the Nuclear Regulatory Commission (NRC) to perform a probabilistic risk assessment (PRA). Although the PRA study is being performed for a specific nuclear plant, the PRA study is also intended to provide a framework for a general methodology that could also be applied to other dry cask systems at other nuclear plants. The dry cask system consists of a transfer cask, used for handling and moving the multi-purpose canister OLIIpC that contains the fuel, and a storage cask, used to store the MPC and fuel on a concrete pad at the site. This paper describes the analyses of the casks for two loading events. The first loading consists of dropping the transfer cask while it is lowered by a crane to a concrete floor at ground elevation. The second loading consists of dropping the storage cask while it is being transferred to the concrete storage pad outdoors. Three dimensional finite element models of the transfer cask and storage cask, containing the MPC and fuel, were utilized to perform the drop analyses. These models were combined with finite element models of the target structures being impacted. The transfer cask drop analyses considered various drop heights for the cask impacting the reinforced concrete floor at ground level. The finite element model of the target included a section of the concrete floor and concrete wall supporting the floor. The storage cask drop analyses evaluated a 30.5 cm (12 in.) drop of the cask impacting three different surfaces: reinforced concrete, asphalt, and gravel.

  7. The effect of blade tenderization on the palatability and retail caselife of beef steaks

    E-Print Network [OSTI]

    Huerta, Nelson Orlando

    1976-01-01T23:59:59.000Z

    of the Student Newman-Keuls' test. (Steel and Torrie, 1960). E i tie. S * t*pra ds (ZMPS168) ad 7 bottom rounds (IMPS 171B) were wrapped in polyvinyl chloride (PVC) film (Goodyear "Prime Wrap" ) and stored for 23-25 days at 3-4 C in order to develop slime... steak. All steaks in this experiment were placed in individual styroi'oam trays, overwrapped with PVC film (Goodyear "Prime 25 Wrap" ) and placed under simulated retail caselife conditions (1-3 C under 90 ft-C of incandescent light). A trained 2...

  8. Estimation of the reliability of space nuclear power systems by probabilistic risk assessment techniques

    E-Print Network [OSTI]

    Gutner, Sophie Isabelle

    1996-01-01T23:59:59.000Z

    of electrons between the emitter and collector, a positively charged ion plasma is injected between the elecuodes. In the SEHPTR's TI assemblies (detailed in Chapter VI), the plasma is cesium. The temperatures of the emitter and collector are 1925K and 1000K... and the contamination of land, air, water, and the food supply. It also provides insights into the relative effectiveness of emergency response planning. PRA permits one to identify components and system failure modes and to obtain a clear understanding of how one...

  9. Surveillance test interval optimization

    SciTech Connect (OSTI)

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

    1995-12-31T23:59:59.000Z

    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.

  10. Generating optimal states for a homodyne Bell test

    E-Print Network [OSTI]

    Sonja Daffer; Peter L. Knight

    2005-04-12T23:59:59.000Z

    We present a protocol that produces a conditionally prepared state that can be used for a Bell test based on homodyne detection. Based on the results of Munro [PRA 1999], the state is near-optimal for Bell-inequality violations based on quadrature-phase homodyne measurements that use correlated photon-number states. The scheme utilizes the Gaussian entanglement distillation protocol of Eisert et. al. [Annals of Phys. 2004] and uses only beam splitters and photodetection to conditionally prepare a non-Gaussian state from a source of two-mode squeezed states with low squeezing parameter, permitting a loophole-free test of Bell inequalities.

  11. Reverse osmosis desalination with osmotic polyelectrolyte intermediate 

    E-Print Network [OSTI]

    McConnell, Thomas Theodore

    1967-01-01T23:59:59.000Z

    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...

  12. Baticum! Curso avançado de português brasileiro, língua estrangeira, a partir de textos da MPB

    E-Print Network [OSTI]

    Simoes, Antonio Roberto Monteiro

    2012-12-01T23:59:59.000Z

    : Vão a.bri.r o .ba.r a.ma.nhã. (enlace em ambos os casos) Vão a.bri’ o .ba.r a.ma.nhã. (enlace com o “r” não ocorre quando há omissão do r no infinitivo (abri); enlace se mantém em “bar”) Para isso é bom usar raio laser. (os rs são pronunciados...) P’ra isso é bom usá raio lêisi. (os rs são omitidos; laser soa como a palavra inglesa, lazy) Essas variantes não devem confundir o estudante. Basta optar por uma da pronúncias e tentar ser consistente. Em caso de dúvida, sugerimos...

  13. Review of Quantitative Software Reliability Methods

    SciTech Connect (OSTI)

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

    2010-09-17T23:59:59.000Z

    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.

  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)

    1994-08-01T23:59:59.000Z

    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. Subsystem fragility: Seismic Safety Margins Research Program (Phase I)

    SciTech Connect (OSTI)

    Kennedy, R. P.; Campbell, R. D.; Hardy, G.; Banon, H.

    1981-10-01T23:59:59.000Z

    Seismic fragility levels of safety related equipment are developed for use in a seismic oriented Probabilistic Risk Assessment (PRA) being conducted as part of the Seismic Safety Margins Research Program (SSMRP). The Zion Nuclear Power Plant is being utilized as a reference plant and fragility descriptions are developed for specific and generic safety related equipment groups in Zion. Both equipment fragilities and equipment responses are defined in probabilistic terms to be used as input to the SSMRP event tree/fault tree models of the Zion systems. 65 refs., 14 figs., 11 tabs.

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

    E-Print Network [OSTI]

    Gellner, David N; LeVine, Sarah

    2007-01-01T23:59:59.000Z

    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...

  17. Adaptive Sampling using Support Vector Machines

    SciTech Connect (OSTI)

    D. Mandelli; C. Smith

    2012-11-01T23:59:59.000Z

    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.

  18. Demonstration of deterministic and high fidelity squeezing of quantum information

    E-Print Network [OSTI]

    Jun-ichi Yoshikawa; Toshiki Hayashi; Takayuki Akiyama; Nobuyuki Takei; Alexander Huck; Ulrik L. Andersen; Akira Furusawa

    2007-11-09T23:59:59.000Z

    By employing at recent proposal (R. Filip, P. Marek and U.L. Andersen, Phys. Rev. A {\\bf 71}, 042308 (2005) \\cite{Filip05.pra}), we experimentally demonstrate a universal, deterministic and high-fidelity squeezing transformation of an optical field. It relies only on linear optics, homodyne detection, feedforward and an ancillary squeezed vacuum state, thus direct interaction between a strong pump and the quantum state is circumvented. We demonstrate three different squeezing levels for a coherent state input. This scheme is highly suitable for the fault-tolerant squeezing transformation in a continuous variable quantum computer.

  19. Use of neural networks to correlate enzymatic hydrolysis with biomass properties

    E-Print Network [OSTI]

    Narayan, Ramasubramanian

    2001-01-01T23:59:59.000Z

    % 38 40 42 prediction interval. 44 Page 3. 9 Correlation between 1-h total sugar conversions and L/G, A/G, L/X, A/X, Xo, and CrI. The dotted lines describe 95% prediction interval . . 45 3. 10 Correlation between 3-d total sugar conversions... Conversion (%) 100 op 0 0 8 ~ rs egest ~ 0-, a t e 0 pY ~ ~ ~ e ~ ~ e gs & ~ ~ ~ A R* ~ 9. 8N7 ~ ~ ? ? 95% PraItiegon t~ N Catculraad ~ Total Sugar Conversion (%1 Figure 1. 4 Correlation between total sugar conversions and L/G, A/G, L...

  20. SAPHIRE 8 Software Project Plan

    SciTech Connect (OSTI)

    Curtis L.Smith; Ted S. Wood

    2010-03-01T23:59:59.000Z

    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.

  1. Risk Management for Sodium Fast Reactors.

    SciTech Connect (OSTI)

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

    2015-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Powell, Danny H [ORNL] [ORNL; Elwood Jr, Robert H [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

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

  3. Evaluation of severe accident risk during mid-loop operation at Surry unit-1

    SciTech Connect (OSTI)

    Mubayi, V.; Jo, J.; Lin, C.C.; Neymotin, L.; Pratt, W.T.

    1996-06-01T23:59:59.000Z

    In the past most probabilistic risk assessments (PRAs) of severe accidents in nuclear power plants have considered initiating events which could potentially lead to core damage and containment failure during normal full power operation. However, recent studies and operational experience during periods while plants were shutdown for maintenance or refueling indicated that potential accidents initiated during low power operation or shutdown conditions could also potentially become important contributors to risk. In 1989, the Nuclear Regulatory Commission (NRC) began an extensive program to assess the risk during low power and shutdown operation. Two plants, Surry (a pressurized water reactor, PWR) and Grand Gulf (a boiling water reactor ,BWR) were selected as the plants to be studied.This paper describes an analysis of accident progression and offsite consequences (level 3 PRA) carried out for the Surry plant. The focus of the level 3 PRA was on mid-loop operation, which is a plant operational state (POS) that can occur while the plant is shutdown for maintenance or refueling. Mid-loop refers to a configuration when the reactor coolant system is lowered to the mid-plane of the hot leg to allow essential maintenance to be performed. This operational state was selected after an initial coarse screening study indicated that reduced inventory during mid-loop operation could pose higher risk than other POSs.

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

    SciTech Connect (OSTI)

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

    1985-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2008-08-01T23:59:59.000Z

    The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer running the Microsoft Windows? operating system. SAPHIRE is primarily funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer and tester. Using the SAPHIRE analysis engine and relational database is a complementary program called GEM. GEM has been designed to simplify using existing PRA analysis for activities such as the NRC’s Accident Sequence Precursor program. In this report, the theoretical framework behind GEM-type calculations are discussed in addition to providing guidance and examples for performing evaluations when using the GEM software. As part of this analysis framework, the two types of GEM analysis are outlined, specifically initiating event (where an initiator occurs) and condition (where a component is failed for some length of time) assessments.

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

    SciTech Connect (OSTI)

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

    1994-08-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Vesely, W.E.; Rezos, J.T. [Science Applications International Corp., Dublin, OH (United States)

    1995-09-01T23:59:59.000Z

    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.

  8. An evaluation of the B&W Owners Group BAW-10182 topical report: Justification for increasing the engineered safety features actuation system on-line test intervals. Technical evaluation report

    SciTech Connect (OSTI)

    Smith, C.L.; Hansen, J.L.

    1993-09-01T23:59:59.000Z

    This Technical Evaluation Report provides an evaluation of the Babcock and Wilcox Owners Group (B&WOG) Technical Specifications Committee Topical Report BAW-10182, entitled, ``Justification for Increasing Engineered Safety Features Actuation System (ESFAS) On-Line Test Intervals.`` This evaluation was performed by the Idaho National Engineering Laboratory in support of the Nuclear Regulatory Commission. The BAW-10182 report presents justification for the extension of on-line test intervals from the existing one-month interval to a three-month interval for the ESFAS system. In the BAW-10182 report, the B&WOG stated that ``{hor_ellipsis}the B&WOG proposes to increase the ESFAS test interval from one to three months and concludes that the effect on plant risk is insignificant.`` The proposed extension was based upon risk-based [i.e., probabilistic risk assessment (PRA)] methods such as reliability block diagrams, uncertainty analyses, and time-dependent system availability analyses. This use of PRA methods requires a detailed evaluation to determine whether the chosen methods and their application are valid in the context of the proposed test interval extension. The results of the evaluation agreed that the effect on plant risk is small if the ESFAS test interval is extended to three months for the ESFAS designs that were evaluated.

  9. 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

    1996-03-01T23:59:59.000Z

    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.

  10. Using population risk assessment as a basis for administrative decisions related to storage of irradiated nuclear fuel

    SciTech Connect (OSTI)

    Droppo, James G. [Pacific Northwest National Laboratory, PO Box 999, Richland WA 99352 (United States); Eremenko, V.A. [International Knowledge Bridge LLC (Russian Federation); Linde, J. [Association on Computer Technology and Informational Systems - ACTIS (Russian Federation); Shilova, E. [Moscow Institute of International Economic Relations, 76, Vernadsky av. 119454 Moscow (Russian Federation)

    2007-07-01T23:59:59.000Z

    Available in abstract form only. Full text of publication follows: Optimization of safety related decisions by local authorities could be improved using information on potential risks to a regional population. A joint Russia-US effort in 2001-2002 modeled potential population health risks for a proposed nuclear waste storage facility in northern Russia. Conducting such an assessment in addition to the standard PRA is proposed as an innovation in Russia aimed at better meeting the needs of local decision makers. This case-study analysis was conducted for the proposed facility to provide insights into potential population health risks. In the case study results, the background population risks from radiation accident exposures were very low compared to risks from chemical background exposures - an unexpected outcome for those that perceive any nuclear facility as very hazardous to the local population. The paper notes that rather than requiring a proposed low-risk facility for hazardous materials be made even safer, these results give the local authority the option of proposing a trade-off of having a major unrelated regional risks mitigated. The results show the value of conducting a population risk assessment in addition to a facility-oriented PRA as a means of better defining the potential impacts. (authors)

  11. Component Fragility Research Program: Phase 1 component prioritization

    SciTech Connect (OSTI)

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

    1987-06-01T23:59:59.000Z

    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.

  12. Dynamical systems probabilistic risk assessment.

    SciTech Connect (OSTI)

    Denman, Matthew R.; Ames, Arlo Leroy

    2014-03-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Jeffrey C. JOe; Ronald L. Boring

    2014-06-01T23:59:59.000Z

    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.

  14. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA)); Sprung, J.L.; Ritchie, L.T.; Jow, Hong-Nian (Sandia National Labs., Albuquerque, NM (USA))

    1990-02-01T23:59:59.000Z

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

  15. MELCOR Accident Consequence Code System (MACCS)

    SciTech Connect (OSTI)

    Jow, H.N.; Sprung, J.L.; Ritchie, L.T. (Sandia National Labs., Albuquerque, NM (USA)); Rollstin, J.A. (GRAM, Inc., Albuquerque, NM (USA)); Chanin, D.I. (Technadyne Engineering Consultants, Inc., Albuquerque, NM (USA))

    1990-02-01T23:59:59.000Z

    This report describes the MACCS computer code. The purpose of this code is to simulate the impact of severe accidents at nuclear power plants on the surrounding environment. MACCS has been developed for the US Nuclear Regulatory Commission to replace the previously used CRAC2 code, and it incorporates many improvements in modeling flexibility in comparison to CRAC2. The principal phenomena considered in MACCS are atmospheric transport, mitigative actions based on dose projection, dose accumulation by a number of pathways including food and water ingestion, early and latent health effects, and economic costs. The MACCS code can be used for a variety of applications. These include (1) probabilistic risk assessment (PRA) of nuclear power plants and other nuclear facilities, (2) sensitivity studies to gain a better understanding of the parameters important to PRA, and (3) cost-benefit analysis. This report is composed of three volumes. Volume I, the User's Guide, describes the input data requirements of the MACCS code and provides directions for its use as illustrated by three sample problems. Volume II, the Model Description, describes the underlying models that are implemented in the code, and Volume III, the Programmer's Reference Manual, describes the code's structure and database management. 59 refs., 14 figs., 15 tabs.

  16. The Cognitive Environment Simulation as a tool for modeling human performance and reliability

    SciTech Connect (OSTI)

    Woods, D.D. (Ohio State Univ., Columbus, OH (USA). Cognitive Systems Engineering Lab.); Pople, H.E. Jr. (Seer Systems (USA)); Roth, E.M. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Science and Technology Center)

    1990-06-01T23:59:59.000Z

    The US Nuclear Regulatory Commission is sponsoring a program to develop improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. A tool called Cognitive Environment Simulation (CES) was developed for simulating how people form intentions to act in NPP emergencies. CES provides an analytic tool for exploring plausible human responses in emergency situations. In addition a methodology called Cognitive Reliability Assessment Technique (CREATE) was developed that describes how CES can be used to provide input to human reliability analyses (HRA) in probabilistic risk assessment (PRA) studies. This report describes the results of three activities that were performed to evaluate CES/CREATE: (1) A technical review was conducted by a panel of experts in cognitive modeling, PRA and HRA; (2) CES was exercised on steam generator tube rupture incidents for which data on operator performance exist; (3) a workshop with HRA practitioners was held to analyze a worked example'' of the CREATE methodology. The results of all three evaluations indicate that CES/CREATE is a promising approach for modeling intention formation. Volume 1 provides a summary of the results. This document, Volume 2, provides details on the three evaluations, including the CES computer outputs for the tube rupture events. 18 refs., 9 figs., 5 tabs.

  17. The Cognitive Environment Simulation as a tool for modeling human performance and reliability

    SciTech Connect (OSTI)

    Woods, D.D. (Ohio State Univ., Columbus, OH (USA). Cognitive Systems Engineering Lab.); Pople, H.E. Jr. (Seer Systems (USA)); Roth, E.M. (Westinghouse Electric Corp., Pittsburgh, PA (USA). Science and Technology Center)

    1990-06-01T23:59:59.000Z

    The US Nuclear Regulatory Commission is sponsoring a program to develop improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. A tool called Cognitive Environment Simulation (CES) was developed for simulating how people form intentions to act in NPP emergencies. CES provides an analytic tool for exploring plausible human response in emergency situations. In addition a methodology called Cognitive Reliability Assessment Technique (CREATE) was developed that describes how CES can be used to provide input to human reliability analyses (HRA) in probabilistic risk assessment (PRA) studies. This report describes the results of three activities that were performed to evaluate CES/CREATE: (1) A technical review was conducted by a panel of experts in cognitive modeling, PRA and HRA; (2) CES was exercised on steam generator tube rupture incidents for which data on operator performance exist; (3) a workshop with HRA practitioners was held to analyze a worked example'' of the CREATE methodology. The results of all three evaluations indicate that CES/CREATE is a promising approach for modeling intention formation. This document, Volume 1 provides a summary of the results. Volume 2 provides details on three evaluations, including the CES computer outputs for the tube rupture events. 14 refs., 3 figs.

  18. The change in risk due to upgrade to a digital reactor protection system

    SciTech Connect (OSTI)

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

    1994-10-01T23:59:59.000Z

    The United States Nuclear Regulatory Commission (NRC) is sponsoring the Idaho National Engineering Laboratory (INEL) to study the risk impact of the upgrade of commercial nuclear power plants (NPPs) from analog to digital instrumentation. It was decided to use a limited probabilistic risk assessment (PRA) methodology to assess the risk impact due to the implementation of digital control systems. The human reliability analysis (HRA)/PRA analysis was to be a pre- and post-upgrade comparison to determine the change in risk. A candidate NPP was selected as the initial step in the analysis process. The selection process entailed reviewing the literature to identify NPPs that had undergone a modification in which a digital control system (DCS) replaced an analog control system. Preference was given to plants where the DCS controlled safety systems. Several NPPs were considered for the candidate plant. The NPP was selected primarily because: (1) it had replaced its existing analog control system with the Westinghouse Eagle-21 Process Protection System (Eagle-21); (2) it had several years experience with the DCS; and (3) the NPP`s management demonstrated a willingness to support the project.

  19. ISSUES ASSOCIATED WITH PROBABILISTIC FAILURE MODELING OF DIGITAL SYSTEMS

    SciTech Connect (OSTI)

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

    2004-09-19T23:59:59.000Z

    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.

  20. 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.

    1992-11-01T23:59:59.000Z

    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.

  1. Preliminary Hazards Analysis Plasma Hearth Process

    SciTech Connect (OSTI)

    Aycock, M.; Coordes, D.; Russell, J.; TenBrook, W.; Yimbo, P. [Science Applications International Corp., Pleasanton, CA (United States)] [Science Applications International Corp., Pleasanton, CA (United States)

    1993-11-01T23:59:59.000Z

    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.

  2. 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.

    2013-04-04T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Holcomb, David Eugene [ORNL

    2015-01-01T23:59:59.000Z

    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.

  4. 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

    2013-10-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2011-01-01T23:59:59.000Z

    A nuclear material control and accountability (MC&A) system effectiveness tool (MSET) has been developed in the United States for use in evaluating material protection, control, and accountability (MPC&A) systems in nuclear facilities. The project was commissioned by the National Nuclear Security Administration's Office of International Material Protection and Cooperation. MSET was developed by personnel with experience spanning more than six decades in both the U.S. and international nuclear programs and with experience in probabilistic risk assessment (PRA) in the nuclear power industry. MSET offers significant potential benefits for improving nuclear safeguards and security in any nation with a nuclear program. MSET provides a design basis for developing an MC&A system at a nuclear facility that functions to protect against insider theft or diversion of nuclear materials. MSET analyzes the system and identifies several risk importance factors that show where sustainability is essential for optimal performance and where performance degradation has the greatest impact on total system risk. MSET contains five major components: (1) A functional model that shows how to design, build, implement, and operate a robust nuclear MC&A system (2) A fault tree of the operating MC&A system that adapts PRA methodology to analyze system effectiveness and give a relative risk of failure assessment of the system (3) A questionnaire used to document the facility's current MPC&A system (provides data to evaluate the quality of the system and the level of performance of each basic task performed throughout the material balance area [MBA]) (4) A formal process of applying expert judgment to convert the facility questionnaire data into numeric values representing the performance level of each basic event for use in the fault tree risk assessment calculations (5) PRA software that performs the fault tree risk assessment calculations and produces risk importance factor reports on the facility's MC&A (software widely used in the aerospace, chemical, and nuclear power industries) MSET was peer reviewed in 2007 and validated in 2008 by benchmark testing at the Idaho National Laboratory in the United States. The MSET documents were translated into Russian and provided to Rosatom in July of 2008, and MSET is currently being evaluated for potential application in Russian Nuclear Facilities.

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

    SciTech Connect (OSTI)

    Lewis, Stuart R. (Electric Power Research Institute, Charlotte, NC); Cooper, Susan E. (U.S. Nuclear Regulatory Commission, Rockville, MD); Najafi, Bijan (SAIC, Campbell, CA); Collins, Erin (SAIC, Campbell, CA); Hannaman, Bill (SAIC, Campbell, CA); Kohlhepp, Kaydee (Scientech, Tukwila, WA); Grobbelaar, Jan (Scientech, Tukwila, WA); Hill, Kendra (U.S. Nuclear Regulatory Commission, Rockville, MD); Hendrickson, Stacey M. Langfitt; Forester, John Alan; Julius, Jeff (Scientech, Tukwila, WA)

    2010-03-01T23:59:59.000Z

    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.

  7. Performing Probabilistic Risk Assessment Through RAVEN

    SciTech Connect (OSTI)

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

    2013-06-01T23:59:59.000Z

    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

  8. Entangling Power of Permutations

    E-Print Network [OSTI]

    Lieven Clarisse; Sibasish Ghosh; Simone Severini; Anthony Sudbery

    2005-04-11T23:59:59.000Z

    The notion of entangling power of unitary matrices was introduced by Zanardi, Zalka and Faoro [PRA, 62, 030301]. We study the entangling power of permutations, given in terms of a combinatorial formula. We show that the permutation matrices with zero entangling power are, up to local unitaries, the identity and the swap. We construct the permutations with the minimum nonzero entangling power for every dimension. With the use of orthogonal latin squares, we construct the permutations with the maximum entangling power for every dimension. Moreover, we show that the value obtained is maximum over all unitaries of the same dimension, with possible exception for 36. Our result enables us to construct generic examples of 4-qudits maximally entangled states for all dimensions except for 2 and 6. We numerically classify, according to their entangling power, the permutation matrices of dimension 4 and 9, and we give some estimates for higher dimensions.

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

    SciTech Connect (OSTI)

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

    2013-10-01T23:59:59.000Z

    A recent trend in the nuclear power engineering field is the implementation of heavily computational and time consuming algorithms and codes for both design and safety analysis. In particular, the new generation of system analysis codes aim to embrace several phenomena such as thermo-hydraulic, structural behavior, and system dynamics, as well as uncertainty quantification and sensitivity analyses. The use of dynamic probabilistic risk assessment (PRA) methodologies allows a systematic approach to uncertainty quantification. Dynamic methodologies in PRA account for possible coupling between triggered or stochastic events through explicit consideration of the time element in system evolution, often through the use of dynamic system models (simulators). They are usually needed when the system has more than one failure mode, control loops, and/or hardware/process/software/human interaction. Dynamic methodologies are also capable of modeling the consequences of epistemic and aleatory uncertainties. The Monte-Carlo (MC) and the Dynamic Event Tree (DET) approaches belong to this new class of dynamic PRA methodologies. The major challenges in using MC and DET methodologies (as well as other dynamic methodologies) are the heavier computational and memory requirements compared to the classical ET analysis. This is due to the fact that each branch generated can contain time evolutions of a large number of variables (about 50,000 data channels are typically present in RELAP) and a large number of scenarios can be generated from a single initiating event (possibly on the order of hundreds or even thousands). Such large amounts of information are usually very difficult to organize in order to identify the main trends in scenario evolutions and the main risk contributors for each initiating event. This report aims to improve Dynamic PRA methodologies by tackling the two challenges mentioned above using: 1) adaptive sampling techniques to reduce computational cost of the analysis and 2) topology-based methodologies to interactively visualize multidimensional data and extract risk-informed insights. Regarding item 1) we employ learning algorithms that aim to infer/predict simulation outcome and decide the coordinate in the input space of the next sample that maximize the amount of information that can be gained from it. Such methodologies can be used to both explore and exploit the input space. The later one is especially used for safety analysis scopes to focus samples along the limit surface, i.e. the boundaries in the input space between system failure and system success. Regarding item 2) we present a software tool that is designed to analyze multi-dimensional data. 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.

  10. Seismic fragility test of a 6-inch diameter pipe system

    SciTech Connect (OSTI)

    Chen, W. P.; Onesto, A. T.; DeVita, V.

    1987-02-01T23:59:59.000Z

    This report contains the test results and assessments of seismic fragility tests performed on a 6-inch diameter piping system. The test was funded by the US Nuclear Regulatory Commission (NRC) and conducted by ETEC. The objective of the test was to investigate the ability of a representative nuclear piping system to withstand high level dynamic seismic and other loadings. Levels of loadings achieved during seismic testing were 20 to 30 times larger than normal elastic design evaluations to ASME Level D limits would permit. Based on failure data obtained during seismic and other dynamic testing, it was concluded that nuclear piping systems are inherently able to withstand much larger dynamic seismic loadings than permitted by current design practice criteria or predicted by the probabilistic risk assessment (PRA) methods and several proposed nonlinear methods of failure analysis.

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

    SciTech Connect (OSTI)

    Swain, A D; Guttmann, H E

    1983-08-01T23:59:59.000Z

    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.

  12. Spatial search using the discrete time quantum walk

    E-Print Network [OSTI]

    Neil B. Lovett; Matthew Everitt; Matthew Trevers; Daniel Mosby; Dan Stockton; Viv Kendon

    2010-10-22T23:59:59.000Z

    We study the quantum walk search algorithm of Shenvi, Kempe and Whaley [PRA 67 052307 (2003)] on data structures of one to two spatial dimensions, on which the algorithm is thought to be less efficient than in three or more spatial dimensions. Our aim is to understand why the quantum algorithm is dimension dependent whereas the best classical algorithm is not, and to show in more detail how the efficiency of the quantum algorithm varies with spatial dimension or accessibility of the data. Our numerical results agree with the expected scaling in 2D of $O(\\sqrt{N \\log N})$, and show how the prefactors display significant dependence on both the degree and symmetry of the graph. Specifically, we see, as expected, the prefactor of the time complexity dropping as the degree (connectivity) of the structure is increased.

  13. The quantum walk search algorithm: Factors affecting efficiency

    E-Print Network [OSTI]

    Neil B. Lovett; Matthew Everitt; Robert M. Heath; Viv Kendon

    2011-10-21T23:59:59.000Z

    We numerically study the quantum walk search algorithm of Shenvi, Kempe and Whaley [PRA \\textbf{67} 052307] and the factors which affect its efficiency in finding an individual state from an unsorted set. Previous work has focused purely on the effects of the dimensionality of the dataset to be searched. Here, we consider the effects of interpolating between dimensions, connectivity of the dataset, and the possibility of disorder in the underlying substrate: all these factors affect the efficiency of the search algorithm. We show that, as well as the strong dependence on the spatial dimension of the structure to be searched, there are also secondary dependencies on the connectivity and symmetry of the lattice, with greater connectivity providing a more efficient algorithm. In addition, we also show that the algorithm can tolerate a non-trivial level of disorder in the underlying substrate.

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

    SciTech Connect (OSTI)

    Lois, Erasmia (US Nuclear Regulatory Commission); Forester, John Alan; Tran, Tuan Q. (Idaho National Laboratory, Idaho Falls, ID); Hendrickson, Stacey M. Langfitt; Boring, Ronald L. (Idaho National Laboratory, Idaho Falls, ID)

    2008-04-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    Auflick, J.L.

    1999-08-12T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1998-05-27T23:59:59.000Z

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

  17. Diagnostico de Prenez en Ganado Vacuno.

    E-Print Network [OSTI]

    Sorensen, A. M. Jr.; Beverly, J. R.; Arias, A. A.

    1975-01-01T23:59:59.000Z

    del Texas Agricultural Extension Service cle su contlaclo. Para c:~lcul:~r 10s costos de operacitin, tome un precio especifico de 27 centavos y coloque el punto en el cual ni g;ma ni pierde dinero (I~reak-even point) con ese precio en la tal~la 1.... Si 1~1ce esto, se requiere destetar un 90 poi- ciento de 10s I,ecerros que pesan 400 1il11-as p:~ra Ilegar a1 punto en el cual ni gana ni pierde dinero. Si sGlo desteta un 60 por ciento de 10s I~ecerros que pesan 400 libras, entonces el costo...

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

    SciTech Connect (OSTI)

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

    2013-05-01T23:59:59.000Z

    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.

  19. Fifty years of progress in reactor safety

    SciTech Connect (OSTI)

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

    1992-01-01T23:59:59.000Z

    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.

  20. Few-Body Bound States of Dipole-Dipole Interacting Rydberg Atoms

    E-Print Network [OSTI]

    Martin Kiffner; Mingxia Huo; Wenhui Li; Dieter Jaksch

    2014-07-08T23:59:59.000Z

    We show that the resonant dipole-dipole interaction can give rise to bound states between two and three Rydberg atoms with non-overlapping electron clouds. The dimer and trimer states arise from avoided level crossings between states converging to different fine structure manifolds in the limit of separated atoms. We analyze the angular dependence of the potential wells, characterize the quantum dynamics in these potentials and discuss methods for their production and detection. Typical distances between the atoms are of the order of several micrometers which can be resolved in state-of-the-art experiments. The potential depths and typical oscillation frequencies are about one order of magnitude larger as compared to the dimer and trimer states investigated in [PRA $\\textbf{86}$ 031401(R) (2012)] and [PRL $\\textbf{111}$ 233003 (2014)], respectively. We find that the dimer and trimer molecules can be aligned with respect to the axis of a weak electric field.

  1. Simplified Expert Elicitation Procedure for Risk Assessment of Operating Events

    SciTech Connect (OSTI)

    Ronald L. Boring; David Gertman; Jeffrey Joe; Julie Marble; William Galyean; Larry Blackwood; Harold Blackman

    2005-06-01T23:59:59.000Z

    This report describes a simplified, tractable, and usable procedure within the US Nuclear Regulator Commission (NRC) for seeking expert opinion and judgment. The NRC has increased efforts to document the reliability and risk of nuclear power plants (NPPs) through Probabilistic Risk Assessment (PRA) and Human Reliability Analysis (HRA) models. The Significance Determination Process (SDP) and Accident Sequence Precursor (ASP) programs at the NRC utilize expert judgment on the probability of failure, human error, and the operability of equipment in cases where otherwise insufficient operational data exist to make meaningful estimates. In the past, the SDP and ASP programs informally sought the opinion of experts inside and outside the NRC. This document represents a formal, documented procedure to take the place of informal expert elicitation. The procedures outlined in this report follow existing formal expert elicitation methodologies, but are streamlined as appropriate to the degree of accuracy required and the schedule for producing SDP and ASP analyses.

  2. Safety research programs sponsored by Office of Nuclear Regulatory Research: Quarterly progress report, July 1-September 30, 1986

    SciTech Connect (OSTI)

    Bari, R.A.; Bezler, P.; Boccio, J.L.; Ginsberg, T.; Greene, G.A.; Guppy, J.G.; Hall, R.E.; Hofmayer, C.H.; Khatib-Rahbar, H.; Luckas, W.J. Jr.

    1987-03-01T23:59:59.000Z

    This progress report will describe current activities and technical progress in the programs at Brookhaven National Laboratory sponsored by the Division of Accident Evaluation, Division of Engineering Technology, and Division of Risk Analysis and Operations of the US Nuclear Regulatory Commission, Office of Nuclear Regulatory Research. The projects reported are the following: High Temperature Reactor Research, SSC code improvements, Thermal-Hydraulic Reactor Safety Experiments, Thermodynamic Core-Concrete Interaction Experiments and Analysis, Plant Analyzer, Code Assessment and Application, Code Maintenance (RAMONA-3B), MELCOR Verification and Benchmarking, Source Term Code Package Verification and Benchmarking, Uncertainty Analysis of the Source Term; Stress Corrosion Cracking of PWR Steam Generator Tubing, Soil-Structure Interaction Evaluation and Structural Benchmarks, Identification of Age Related Failure Modes; Application of HRA/PRA Results to Support Resolution of Generic Safety Issues Involving Human Performance, Protective Action Decisionmaking, Rebaseling of Risk for Zion, Containment Performance Design Objective, and Operational Safety Reliability Research.

  3. 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))

    1989-08-01T23:59:59.000Z

    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.

  4. MELCOR technical assessment at SNL

    SciTech Connect (OSTI)

    Kmetyk, L.N.; Tautges, T.J.

    1993-12-01T23:59:59.000Z

    MELCOR is a fully integrated, engineering-level computer code that models the progression of severe accidents in light water reactor (LWR) nuclear power plants, which is being developed at Sandia National Laboratories for the US Nuclear Regulatory Commission (US-NRC). 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 (PRWs). The MELCOR computer code has been developed to the point that 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 code 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. This report provides a discussion of this technical assessment.

  5. A pulse processing station

    SciTech Connect (OSTI)

    Morgado, A.M.L.S.; Simoes, J.B.; Landeck, J. [Univ. of Coimbra (Portugal)] [and others

    1996-12-31T23:59:59.000Z

    This is the first of two papers concerning the architecture, circuitry design and performance of a pulse processing system based on a digital signal processor. This multifunction system, implemented as a single PC module, incorporates a high performance 16-bit Pulse Height Analyzer (PHA) a Multichannel Scaler (MCS), a Digital Oscilloscope (DSO) and also a Digital Pulse Processor (DPP). This paper presents the PRA architecture with emphasis on the baseline restorer and peak stretcher circuits. Differential nonlinearities (DNL) are corrected by a new implementation of the sliding scale technique and performance ranges from better than 2% (at 16-bit resolution) up to less than 0.2% for 12-bit operation. The DNL correction technique is assessed for different sliding-scale ranges.

  6. Integrated systems analysis of the PIUS reactor

    SciTech Connect (OSTI)

    Fullwood, F.; Kroeger, P.; Higgins, J. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1993-11-01T23:59:59.000Z

    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.

  7. Reviewing the impact of advanced control room technology

    SciTech Connect (OSTI)

    Wilhelmsen, C.A.; Gertman, D.I.; Ostrom, L.T.; Nelson, W.R.; Galyean, W.J.; Byers, J.C.

    1992-01-01T23:59:59.000Z

    Progress to date on assessing the nature of the expected changes in human performance and risk associated with the introduction of digital control, instrumentation, and display systems is presented. Expected changes include the shift toward more supervisory tasks, development of intervention strategies, and reallocation of function between human and machine. Results are reported in terms of the scope of new technology, human performance issues, and crews experience with digital control systems in a variety of industries petrochemical and aerospace. Plans to conduct a limited Probabilistic Risk Assessment/Human Reliability Assessment (PRA/HRA) comparison between a conventional NUREG-1150 series plant and that same plant retrofit with distributed control and advanced instrumentation and display are also presented. Changes needed to supplement existing HRA modeling methods and quantification techniques are discussed.

  8. Reviewing the impact of advanced control room technology

    SciTech Connect (OSTI)

    Wilhelmsen, C.A.; Gertman, D.I.; Ostrom, L.T.; Nelson, W.R.; Galyean, W.J.; Byers, J.C.

    1992-08-01T23:59:59.000Z

    Progress to date on assessing the nature of the expected changes in human performance and risk associated with the introduction of digital control, instrumentation, and display systems is presented. Expected changes include the shift toward more supervisory tasks, development of intervention strategies, and reallocation of function between human and machine. Results are reported in terms of the scope of new technology, human performance issues, and crews experience with digital control systems in a variety of industries petrochemical and aerospace. Plans to conduct a limited Probabilistic Risk Assessment/Human Reliability Assessment (PRA/HRA) comparison between a conventional NUREG-1150 series plant and that same plant retrofit with distributed control and advanced instrumentation and display are also presented. Changes needed to supplement existing HRA modeling methods and quantification techniques are discussed.

  9. Risk management at Argonne National Laboratory

    SciTech Connect (OSTI)

    Hill, D.J.; Hislop, R.D.

    1994-02-01T23:59:59.000Z

    The only facility at Argonne National Laboratory which is classified as high hazard is EBR-II. A Level I Probabilistic Risk Assessment (PRA), including external events, has been completed for EBR-II. There were several objectives for this project; to provide a quantitative estimate of the risk associated with the operation of EBR-II, to provide a framework for managerial decision-making for the management of risk at the facility, and to provide insights into the nature of the risk of EBR-II that can be applied in the design of future LMRS. Other ANL facilities do not have complete probabilistic assessments. Despite this fact, Risk Management is an essential part of ANL`s approach to safety and operations. Risk management at Argonne National Laboratory is not limited to accelerator or nuclear facilities. It is also an integral part of construction activities. The Advanced Photon Source, a major construction project at the Laboratory, utilizes a variety of risk assessment techniques to identify potential construction loss exposures and to develop measures to eliminate them. Over the past 15 years, in excess of 15,000 pages of regulatory requirements pertaining to environment safety and health have been printed in the Federal Register. Not all of these are applicable to Argonne National Laboratory all of the time, but as a highly-visible, nationally-funded facility, compliance with those that are applicable must be above reproach. Therefore, risk management is also a very important part of construction activities at ANL. This paper will give examples of these activities, such as, the EBR-II PRA, risk-based analyses of Fuel Cycle Facility Safety Systems, reliability studies of the Access Control Interlock System for the Advanced Photon Source and management approaches for controlling risk during the construction activities at APS.

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

    SciTech Connect (OSTI)

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

    1994-06-01T23:59:59.000Z

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

  11. Discussion of comments from a peer review of a technique for human event analysis (ATHEANA)

    SciTech Connect (OSTI)

    Forester, J.A. [Sandia National Labs., Albuquerque, NM (United States); Ramey-Smith, A. [Nuclear Regulatory Commission, Washington, DC (United States); Bley, D.C. [Buttonwood Consulting, Inc. (United States); Kolaczkowski, A.M.; Cooper, S.E. [Science Applications International Corp. (United States); Wreathall, J. [John Wreathall and Company (United States)

    1998-09-01T23:59:59.000Z

    In May of 1998, a technical basis and implementation guidelines document for A Technique for Human Event Analysis (ATHEANA) was issued as a draft report for public comment (NUREG-1624). In conjunction with the release of the draft NUREG, a paper review of the method, its documentation, and the results of an initial test of the method was held over a two-day period in Seattle, Washington, in June of 1998. Four internationally-known and respected experts in human reliability analysis (HRA) were selected to serve as the peer reviewers and were paid for their services. In addition, approximately 20 other individuals with an interest in HRA and ATHEANA also attended the peer review meeting and were invited to provide comments. The peer review team was asked to comment on any aspect of the method or the report in which improvements could be made and to discuss its strengths and weaknesses. All of the reviewers thought the ATEANA method had made significant contributions to the field of PRA/HRA, in particular by addressing the most important open questions and issues in HRA, by attempting to develop an integrated approach, and by developing a framework capable of identifying types of unsafe actions that generally have not been considered using existing methods. The reviewers had many concerns about specific aspects of the methodology and made many recommendations for ways to improve and extend the method, and to make its application more cost effective and useful to PRA in general. Details of the reviewers` comments and the ATHEANA team`s responses to specific criticisms will be discussed.

  12. Results of a nuclear power plant Application of a new technique for human error analysis (ATHEANA)

    SciTech Connect (OSTI)

    Forester, J.A.; Whitehead, D.W.; Kolaczkowski, A.M.; Thompson, C.M.

    1997-10-01T23:59:59.000Z

    A new method to analyze human errors has been demonstrated at a pressurized water reactor (PWR) nuclear power plant. This was the first application of the new method referred to as A Technique for Human Error Analysis (ATHEANA). The main goals of the demonstration were to test the ATHEANA process as described in the frame-of-reference manual and the implementation guideline, test a training package developed for the method, test the hypothesis that plant operators and trainers have significant insight into the error-forcing-contexts (EFCs) that can make unsafe actions (UAs) more likely, and to identify ways to improve the method and its documentation. A set of criteria to evaluate the {open_quotes}success{close_quotes} of the ATHEANA method as used in the demonstration was identified. A human reliability analysis (HRA) team was formed that consisted of an expert in probabilistic risk assessment (PRA) with some background in HRA (not ATHEANA) and four personnel from the nuclear power plant. Personnel from the plant included two individuals from their PRA staff and two individuals from their training staff. Both individuals from training are currently licensed operators and one of them was a senior reactor operator {open_quotes}on shift{close_quotes} until a few months before the demonstration. The demonstration was conducted over a 5 month period and was observed by members of the Nuclear Regulatory Commission`s ATHEANA development team, who also served as consultants to the HRA team when necessary. Example results of the demonstration to date, including identified human failure events (HFEs), UAs, and EFCs are discussed. Also addressed is how simulator exercises are used in the ATHEANA demonstration project.

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

    SciTech Connect (OSTI)

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

    1994-06-01T23:59:59.000Z

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

  14. VALIDATION OF NUCLEAR MATERIAL CONTROL AND ACCOUNTABILITY (MC&A) SYSTEM EFFECTIVENESS TOOL (MSET) AT IDAHO NATIONAL LABORATORY (INL)

    SciTech Connect (OSTI)

    Meppen, Bruce; Haga, Roger; Moedl, Kelley; Bean, Tom; Sanders, Jeff; Thom, Mary Alice

    2008-07-01T23:59:59.000Z

    A Nuclear Material Control and Accountability (MC&A) Functional Model has been developed to describe MC&A systems at facilities possessing Category I or II Special Nuclear Material (SNM). Emphasis is on achieving the objectives of 144 “Fundamental Elements” in key areas ranging from categorization of nuclear material to establishment of Material Balance Areas (MBAs), controlling access, performing quality measurements of inventories and transfers, timely reporting all activities, and detecting and investigating anomalies. An MC&A System Effectiveness Tool (MSET), including probabilistic risk assessment (PRA) technology for evaluating MC&A effectiveness and relative risk, has been developed to accompany the Functional Model. The functional model and MSET were introduced at the 48th annual International Nuclear Material Management (INMM) annual meeting in July, 20071,2. A survey/questionnaire is used to accumulate comprehensive data regarding the MC&A elements at a facility. Data is converted from the questionnaire to numerical values using the DELPHI method and exercises are conducted to evaluate the overall effectiveness of an MC&A system. In 2007 a peer review was conducted and a questionnaire was completed for a hypothetical facility and exercises were conducted. In the first quarter of 2008, a questionnaire was completed at Idaho National Laboratory (INL) and MSET exercises were conducted. The experience gained from conducting the MSET exercises at INL helped evaluate the completeness and consistency of the MC&A Functional Model, descriptions of fundamental elements of the MC&A Functional Model, relationship between the MC&A Functional Model and the MC&A PRA tool and usefulness of the MSET questionnaire data collection process.

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

    SciTech Connect (OSTI)

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

    1994-06-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

    J. W. Nielsen; Akira Tokuhiro; Robert Hiromoto

    2014-06-01T23:59:59.000Z

    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.

  17. RAVEN: Dynamic Event Tree Approach Level III Milestone

    SciTech Connect (OSTI)

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

    2013-07-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    1995-01-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    2014-07-30T23:59:59.000Z

    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.

  20. Conversion of Questionnaire Data

    SciTech Connect (OSTI)

    Powell, Danny H [ORNL] [ORNL; Elwood Jr, Robert H [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    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.

  1. 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.

    2008-05-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    Carelli, M.D.; Petrovic, B.

    2004-10-03T23:59:59.000Z

    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.

  3. Quantum heat engines and information

    E-Print Network [OSTI]

    Ye Yeo; Chang Chi Kwong

    2007-08-18T23:59:59.000Z

    Recently, Zhang {\\em et al.} [PRA, {\\bf 75}, 062102 (2007)] extended Kieu's interesting work on the quantum Otto engine [PRL, {\\bf 93}, 140403 (2004)] by considering as working substance a bipartite quantum system $AB$ composed of subsystems $A$ and $B$. In this paper, we express the net work done $W_{AB}$ by such an engine explicitly in terms of the macroscopic bath temperatures and information theoretic quantities associated with the microscopic quantum states of the working substance. This allows us to gain insights into the dependence of positive $W_{AB}$ on the quantum properties of the states. We illustrate with a two-qubit XY chain as the working substance. Inspired by the expression, we propose a plausible formula for the work derivable from the subsystems. We show that there is a critical entanglement beyond which it is impossible to draw positive work locally from the individual subsystems while $W_{AB}$ is positive. This could be another interesting manifestation of quantum nonlocality.

  4. Generic risk insights for Westinghouse and Combustion Engineering pressurized water reactors

    SciTech Connect (OSTI)

    Travis, R.; Taylor, J.; Fresco, A. (Brookhaven National Lab., Upton, NY (USA)); Chung, J. (Nuclear Regulatory Commission, Washington, DC (USA))

    1990-11-01T23:59:59.000Z

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of Westinghouse and Combustion Engineering (CE) pressurized water reactors (PWRs) and apply the insights gained to Westinghouse and Ce plants have not been subjected to a PRA. The available PRAs (five Westinghouse plants and one CE plant) were examined to identify the most probable, i.e., dominant accident sequences at each plant. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eleven sequences met this definition. From these sequences, the most important component failures and human errors that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training maintenance, design review, and inspections.

  5. Generic risk insights for General Electric boiling water reactors

    SciTech Connect (OSTI)

    Travis, R.; Taylor, J. (Brookhaven National Lab., Upton, NY (USA)); Chung, J. (Nuclear Regulatory Commission, Washington, DC (USA). Office of Nuclear Reactor Regulation)

    1991-05-01T23:59:59.000Z

    A methodology has been developed to extract generic risk-based information from probabilistic risk assessments (PRAs) of General Electric boiling water rectors and applying the insights gained to plants that have not been subjected to a PRA. The available risk assessments (six plants) were examined to identify the most probable, i.e., dominant accident sequences at each plants. The goal was to include all sequences which represented at least 80% of core damage frequency. If the same plant specific dominant accident sequence appeared within this boundary in at least two plant PRAs, the sequence was considered to be a representative sequence. Eight sequences met this definition. From these sequences, the most important component failures and human error that contributed to each sequence have been prioritized. Guidance is provided to prioritize the representative sequences and modify selected basic events that have been shown to be sensitive to the plant specific design or operating variations of the contributing PRAs. This risk-based guidance can be used for utility and NRC activities including operator training, maintenance, design review, and inspections. 13 refs., 6 tabs.

  6. 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 [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)] [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)

    2013-07-01T23:59:59.000Z

    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)

  7. Seismic margins and calibration of piping systems

    SciTech Connect (OSTI)

    Shieh, L.C.; Tsai, N.C.; Yang, M.S.; Wong, W.L.

    1985-01-01T23:59:59.000Z

    The Seismic Safety Margins Research Program (SSMRP) is a US Nuclear Regulatory Commission-funded, multiyear program conducted by Lawrence Livermore National Laboratory (LLNL). Its objective is to develop a complete, fully coupled analysis procedure for estimating the risk of earthquake-induced radioactive release from a commercial nuclear power plant and to determine major contributors to the state-of-the-art seismic and systems analysis process and explicitly includes the uncertainties in such a process. The results will be used to improve seismic licensing requirements for nuclear power plants. In Phase I of SSMRP, the overall seismic risk assessment methodology was developed and assembled. The application of this methodology to the seismic PRA (Probabilistic Risk Assessment) at the Zion Nuclear Power Plant has been documented. This report documents the method deriving response factors. The response factors, which relate design calculated responses to best estimate values, were used in the seismic response determination of piping systems for a simplified seismic probablistic risk assessment. 13 references, 31 figures, 25 tables.

  8. Sensitivity of piping seismic responses to input factors

    SciTech Connect (OSTI)

    O'Connell, W.J.

    1985-05-01T23:59:59.000Z

    This report summarizes the sensitivity of peak dynamic seismic responses to input parameters. The responses have been modeled and calculated for the Zion Unit 1 plant as part of a seismic probabilistic risk assessment (PRA) performed by the US NRC Seismic Safety Margins Research Program (SSMRP). The SSMRP was supported by the US NRC Office of Nuclear Regulatory Research. Two sensitivity topics motivated the study. The first is the sensitivity of piping response to the mean value of piping damping. The second is the sensitivity of all the responses to the earthquake and model input parameters including soil, structure and piping parameters; this information is required for another study, the sensitivity of the plant system response (in terms of risk) to these dynamic input parameters and to other input factors. We evaluate the response sensitivities by performing a linear regression analysis (LRA) of the computer code SMACS. With SMACS we have a detailed model of the Zion plant and of the important dynamic processes in the soil, structures and piping systems. The qualitative results change with the location of the individual response. Different responses are in locations where the many potential influences have different effectiveness. The results give an overview of the complexity of the seismic dyanmic response of a plant. Within the diversity trends are evident in the influences of the input variables on the responses.

  9. Risk comparisons based on representative source terms with the NUREG-1150 results

    SciTech Connect (OSTI)

    Mubayi, V.; Davis, R.E.; Hanson, A.L.

    1993-12-01T23:59:59.000Z

    Standardized source terms, based on a specified release of fission products during potential accidents at commercial light water nuclear reactors, have been used for a long time for regulatory purposes. The siting of nuclear power plants, for example, which is governed by Part 100 of the Code of Federal Regulations Title 10, has utilized the source term recommended in TID-14844 supplemented by Regulatory Guides 1.3 and 1.4 and the Standard Review Plan. With the introduction of probabilistic risk assessment (PRA) methods, the source terms became characterized not only by the amount of fission products released, but also by the probability of the release. In the Reactor Safety Study, for example, several categories of source terms, characterized by release severity and probability, were developed for both pressurized and boiling water reactors (PWRs and BWRs). These categories were based on an understanding of the likely paths and associated phenomenology of accident progression following core damage to possible failure of the containment and release to the environment.

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

    SciTech Connect (OSTI)

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

    2013-10-01T23:59:59.000Z

    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.)

  11. An Improved Probabilistic Fracture Mechanics Model for Pressurized Thermal Shock

    SciTech Connect (OSTI)

    Dickson, T.L.

    2001-10-29T23:59:59.000Z

    This paper provides an overview of an improved probabilistic fracture mechanics (PFM) model used for calculating the conditional probabilities of fracture and failure of a reactor pressure vessel (RPV) subjected to pressurized-thermal-shock (PTS) transients. The updated PFM model incorporates several new features: expanded databases for the fracture toughness properties of RPV steels; statistical representations of the fracture toughness databases developed through application of rigorous mathematical procedures; and capability of generating probability distributions for RPV fracture and failure. The updated PFM model was implemented into the FAVOR fracture mechanics program, developed at Oak Ridge National Laboratory as an applications tool for RPV integrity assessment; an example application of that implementation is discussed herein. Applications of the new PFM model are providing essential input to a probabilistic risk assessment (PRA) process that will establish an improved technical basis for re-assessment of current PTS regulations by the US Nuclear Regulatory Commission (NRC). The methodology described herein should be considered preliminary and subject to revision in the PTS re-evaluation process.

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

    1989-08-01T23:59:59.000Z

    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

    2013-04-01T23:59:59.000Z

    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. MELCOR assessment at SNL

    SciTech Connect (OSTI)

    Kmetyk, L. N.

    1992-01-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

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

    2011-10-01T23:59:59.000Z

    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.

  16. Savannah River Reactor Operation: Indices of risk for emergency planning

    SciTech Connect (OSTI)

    O'Kula, K.R.; East, J.M.

    1990-10-01T23:59:59.000Z

    Periodically it is necessary to re-examine the implications of new source terms for neighboring offsite populations as Probabilistic Risk Assessment (PRA) and Severe Accident studies mature, and lead to a better understanding of the progression of hypothetical core melt accidents in the Savannah River Site (SRS) reactors. In this application multiple-system failure, low-frequency events, and consequently higher radiological source terms than from normal operation or design basis accidents (DBAs) are considered. Measures of consequence such as constant dose vs distance, boundary doses, and health effects to close-in populations are usually examined in this context. A set of source terms developed for the Safety Information Document (SID) for support of the Reactor Operation Environmental Impact Statement (EIS) forms the basis for the revised risk evaluation discussed herein. The intent of this review is not to completely substantiate the sufficiency of the current Emergency Planning Zone (EPZ). However, the two principal measures (200-rem red-bone marrow dose vs distance and 300-rem thyroid dose vs distance) for setting an EPZ are considered. Additional dose-at-distance calculations and consideration of DBA doses would be needed to complete a re-evaluation of the current EPZ. These subject areas are not addressed in the current document. Also, this report evaluates the sensitivity of individual risk estimates to the extent of offsite evacuation assumed from a K reactor severe accident and compares these risks to the Draft DOE Safety Guidelines. 14 refs., 8 figs., 4 tabs.

  17. Preliminary risks associated with postulated tritium release from production reactor operation

    SciTech Connect (OSTI)

    O'Kula, K.R.; Horton, W.H.

    1988-01-01T23:59:59.000Z

    The Probabilistic Risk Assessment (PRA) of Savannah River Plant (SRP) reactor operation is assessing the off-site risk due to tritium releases during postulated full or partial loss of heavy water moderator accidents. Other sources of tritium in the reactor are less likely to contribute to off-site risk in non-fuel melting accident scenarios. Preliminary determination of the frequency of average partial moderator loss (including incidents with leaks as small as .5 kg) yields an estimate of /approximately/1 per reactor year. The full moderator loss frequency is conservatively chosen as 5 /times/ 10/sup /minus/3/ per reactor year. Conditional consequences, determined with a version of the MACCS code modified to handle tritium, are found to be insignificant. The 95th percentile individual cancer risk is 4 /times/ 10/sup /minus/8/ per reactor year within 16 km of the release point. The full moderator loss accident contributes about 75% of the evaluated risks. 13 refs., 4 figs., 5 tabs.

  18. Cognitive environment simulation: An artificial intelligence system for human performance assessment: Cognitive reliability analysis technique: (Technical report, May 1986-June 1987)

    SciTech Connect (OSTI)

    Woods, D.D.; Roth, E.M.

    1987-11-01T23:59:59.000Z

    This report documents the results of Phase II of a three phase research program to develop and validate improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. In Phase II a dynamic simulation capability for modeling how people form intentions to act in NPP emergency situations was developed based on techniques from artificial intelligence. This modeling tool, Cognitive Environment Simulation or CES, simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures (e.g., errors of omission, errors of commission, common mode errors), the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person-machine system. The Cognitive Reliability Assessment Technique (or CREATE) was also developed in Phase II to specify how CES can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. 34 refs., 7 figs., 1 tab.

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

    SciTech Connect (OSTI)

    Barriere, M.T.; Luckas, W.J. [Brookhaven National Lab., Upton, NY (United States); Wreathall, J. [Wreathall (John) and Co., Dublin, OH (United States); Cooper, S.E. [Science Applications International Corp., Reston, VA (United States); Bley, D.C. [PLG, Inc., Newport Beach, CA (United States); Ramey-Smith, A. [Nuclear Regulatory Commission, Washington, DC (United States). Div. of Systems Technology

    1995-08-01T23:59:59.000Z

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

  20. Cognitive environment simulation: An artificial intelligence system for human performance assessment: Modeling human intention formation: (Technical report, May 1986-June 1987)

    SciTech Connect (OSTI)

    Woods, D.D.; Roth, E.M.; Pople, H. Jr.

    1987-11-01T23:59:59.000Z

    This report documents the results of Phase II of a three phase research program to develop and validate improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. In Phase II a dynamic simulation capability for modeling how people form intentions to act in NPP emergency situations was developed based on techniques from artificial intelligence. This modeling tool, Cognitive Environment Simulation or CES, simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures (e.g., errors of omission, errors of commission, common mode errors), the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person-machine system. The Cognitive Reliability Assessment Technique (or CREATE) was also developed in Phase II to specify how CES can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. 43 refs., 20 figs., 1 tab.

  1. Cognitive environment simulation: An artificial intelligence system for human performance assessment: Summary and overview: (Technical report, May 1986-June 1987)

    SciTech Connect (OSTI)

    Woods, D.D.; Roth, E.M.; Pople, H. Jr.

    1987-11-01T23:59:59.000Z

    This report documents the results of Phase II of a three phase research program to develop and validate improved methods to model the cognitive behavior of nuclear power plant (NPP) personnel. In Phase II a dynamic simulation capability for modeling how people form intentions to act in NPP emergency situations was developed based on techniques from artificial intelligence. This modeling tool, Cognitive Environment Simulation or CES, simulates the cognitive processes that determine situation assessment and intention formation. It can be used to investigate analytically what situations and factors lead to intention failures, what actions follow from intention failures, the ability to recover from errors or additional machine failures, and the effects of changes in the NPP person-machine system. The Cognitive Reliability Assessment Technique (or CREATE) was also developed in Phase II to specify how CES can be used to enhance the measurement of the human contribution to risk in probabilistic risk assessment (PRA) studies. The results are reported in three self-contained volumes that describe the research from different perspectives. Volume 1 provides an overview of both CES and CREATE. 30 refs., 6 figs.

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

    E-Print Network [OSTI]

    Lopez, Jose M; Avila, Marc

    2015-01-01T23:59:59.000Z

    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...

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

    SciTech Connect (OSTI)

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

    2012-06-01T23:59:59.000Z

    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).

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

    SciTech Connect (OSTI)

    CHENG,L.Y.; LUDEWIG, H.

    2007-06-01T23:59:59.000Z

    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.

  5. Qualitative human reliability analysis for spent fuel handling

    SciTech Connect (OSTI)

    Brewer, J. D. [Sandia National Laboratories, P.O. Box 5800, Albuquerque, NM 87185-0748 (United States); Amico, P. [Science Applications International Corporation (United States); Cooper, S. E. [United Stated Nuclear Regulatory Commission (United States)

    2006-07-01T23:59:59.000Z

    Human reliability analysis (HRA) methods have been developed primarily to provide information for use in probabilistic risk assessments (PRAs) that analyze nuclear power plant (NPP) operations. Given the original emphasis of these methods, it is understandable that many HRAs have not ventured far from NPP control room applications. Despite this historical focus on the control room, there has been growing interest and discussion regarding the application of HRA methods to other NPP activities such as spent fuel handling (SFH) or operations in different types of facilities. One recently developed HRA method, 'A Technique for Human Event Analysis' (ATHEANA) has been proposed as a promising candidate for diverse applications due to its particular approach for systematically uncovering the dynamic, contextual conditions influencing human performance. This paper describes one successful test of this proposition by presenting portions of a recently completed project in which a scoping study was performed to accomplish the following goals: (1) investigate what should be included in a qualitative HRA for spent fuel and cask handling operations; and (2) demonstrate that the ATHEANA HRA technique can be usefully applied to these operations. The preliminary, scoping qualitative HRA examined, in a generic manner, how human performance of SFH and dry cask storage operations (DCSOs) can plausibly lead to radiological consequences that impact the public and the environment. The study involved the performance of typical, qualitative HRA tasks such as collecting relevant information and the preliminary identification of human failure events or unsafe actions, relevant influences (e.g., performance shaping factors, other contextual factors), event scenario development and categorization of human failure event (HFE) scenario groupings. Information from relevant literature sources was augmented with subject matter expert interviews and analysis of an edited video of selected operations. Elements of NUREG-1792, Good Practices for Implementing Human Reliability Analyses (HRA) and NUREG-1624, Rev. 1, Technical Basis and Implementation Guidelines for A Technique for Human Event Analysis (ATHEANA) formed critical parts of the technical basis for the preliminary analysis. Mis-loading of spent fuel into a cask and dropping of a loaded cask were the two human failure event groupings of primary interest, although all human performance aspects of DCSOs were considered to some extent. Of important note is that HRA is typically performed in the context of a plant-specific PRA study. This analysis was performed without the benefit of the context provided by a larger PRA study, nor was it plant specific, and so it investigated only generic HRA issues relevant to SFH. However, the improved understanding of human performance issues provided by the study will likely enhance the ability to carry out a detailed qualitative HRA for a specific NPP at some point in the future. Furthermore, support was obtained regarding the potential for applying ATHEANA beyond NPP settings. This paper provides a description of the process followed during the analysis, a description of the HFE scenario groupings, discussion regarding general human performance vulnerabilities, and a detailed examination of one HFE scenario developed in the study. (authors)

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

    SciTech Connect (OSTI)

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

    2008-10-01T23:59:59.000Z

    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.

  7. 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. [Kidney Research Centre, Ottawa Hospital Research Institute, University of Ottawa, Ontario (Canada)] [Kidney Research Centre, Ottawa Hospital Research Institute, University of Ottawa, Ontario (Canada); Mecawi, André S. [Department of Physiology, Faculty of Medicine of Ribeirão Preto, University of São Paulo (USP), Ribeirão Preto, SP (Brazil)] [Department of Physiology, Faculty of Medicine of Ribeirão Preto, University of São Paulo (USP), Ribeirão Preto, SP (Brazil); Batalhão, Marcelo E.; Carnio, Evelin C. [Department of General and Specialized Nursing, College of Nursing of Ribeirão Preto, USP, São Paulo (Brazil)] [Department of General and Specialized Nursing, College of Nursing of Ribeirão Preto, USP, São Paulo (Brazil); Antunes-Rodrigues, José [Department of Physiology, Faculty of Medicine of Ribeirão Preto, University of São Paulo (USP), Ribeirão Preto, SP (Brazil)] [Department of Physiology, Faculty of Medicine of Ribeirão Preto, University of São Paulo (USP), Ribeirão Preto, SP (Brazil); Queiroz, Regina H. [Department of Clinical, Toxicological and Food Science Analysis, Faculty of Pharmaceutical Sciences, USP, São Paulo (Brazil)] [Department of Clinical, Toxicological and Food Science Analysis, Faculty of Pharmaceutical Sciences, USP, São Paulo (Brazil); Touyz, Rhian M. [Kidney Research Centre, Ottawa Hospital Research Institute, University of Ottawa, Ontario (Canada)] [Kidney Research Centre, Ottawa Hospital Research Institute, University of Ottawa, Ontario (Canada); Tirapelli, Carlos R., E-mail: crtirapelli@eerp.usp.br [Department of Psychiatric Nursing and Human Sciences, Laboratory of Pharmacology, College of Nursing of Ribeirão Preto, USP, Ribeirão Preto, SP (Brazil)

    2012-11-01T23:59:59.000Z

    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.

  8. Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Main report. Volume 6. Part 1

    SciTech Connect (OSTI)

    Jo, J.; Lin, C.C.; Neymotin, L. [Brookhaven National Lab., Upton, NY (United States)] [and others

    1995-05-01T23:59:59.000Z

    During 1989, the Nuclear Regulatory Commission (NRC) initiated an extensive program to carefully examine the potential risks during low power and shutdown operations. The program includes two parallel projects being performed by Brookhaven National Laboratory (BNL) and Sandia National Laboratories (SNL). Two plants, Surry (pressurized water reactor) and Grand Gulf (boiling water reactor), were selected as the plants to be studied. The objectives of the program are to assess the risks of severe accidents initiated during plant operational states other than full power operation and to compare the estimated core damage frequencies, important accident sequences and other qualitative and quantitative results with those accidents initiated during full power operation as assessed in NUREG-1150. The scope of the program includes that of a level-3 PRA. A phased approach was used in the level-1 program. In phase 1 which was completed in Fall 1991, a coarse screening analysis including internal fire and flood was performed for all plant operational states (POSs). The objective of the phase 1 study was to identify potential vulnerable plant configurations, to characterize (on a high, medium, or low basis) the potential core damage accident scenarios, and to provide a foundation for a detailed phase 2 analysis. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed based on the results of the phase 1 study. The objective of the phase 2 study is to perform a detailed analysis of the potential accident scenarios that may occur during mid-loop operation, and compare the results with those of NUREG-1150. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6.

  9. The action characterization matrix: A link between HERA (Human Events Reference for ATHEANA) and ATHEANA (a technique for human error analysis)

    SciTech Connect (OSTI)

    Hahn, H.A.

    1997-12-22T23:59:59.000Z

    The Technique for Human Error Analysis (ATHEANA) is a newly developed human reliability analysis (HRA) methodology that aims to facilitate better representation and integration of human performance into probabilistic risk assessment (PRA) modeling and quantification by analyzing risk-significant operating experience in the context of existing behavior science models. The fundamental premise of ATHEANA is that error-forcing contexts (EFCs), which refer to combinations of equipment/material conditions and performance shaping factors (PSFs), set up or create the conditions under which unsafe actions (UAs) can occur. ATHEANA is being developed in the context of nuclear power plant (NPP) PRAs, and much of the language used to describe the method and provide examples of its application are specific to that industry. Because ATHEANA relies heavily on the analysis of operational events that have already occurred as a mechanism for generating creative thinking about possible EFCs, a database, called the Human Events Reference for ATHEANA (HERA), has been developed to support the methodology. Los Alamos National Laboratory`s (LANL) Human Factors Group has recently joined the ATHEANA project team; LANL is responsible for further developing the database structure and for analyzing additional exemplar operational events for entry into the database. The Action Characterization Matrix (ACM) is conceived as a bridge between the HERA database structure and ATHEANA. Specifically, the ACM allows each unsafe action or human failure event to be characterized according to its representation along each of six different dimensions: system status, initiator status, unsafe action mechanism, information processing stage, equipment/material conditions, and performance shaping factors. This report describes the development of the ACM and provides details on the structure and content of its dimensions.

  10. 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)

    1994-08-01T23:59:59.000Z

    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.

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

    SciTech Connect (OSTI)

    DeKlever, R. [Raytheon Services Nevada, Las Vegas, NV (United States); Verna, B. [Dept. of Energy, Las Vegas, NV (United States)

    1993-12-31T23:59:59.000Z

    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.

  12. Individual plant examination program: Perspectives on reactor safety and plant performance. Part 1: Final summary report; Volume 1

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  13. Individual plant examination program: Perspectives on reactor safety and plant performance. Parts 2--5: Final report; Volume 2

    SciTech Connect (OSTI)

    NONE

    1997-12-01T23:59:59.000Z

    This report provides perspectives gained by reviewing 75 Individual Plant Examination (IPE) submittals pertaining to 108 nuclear power plant units. IPEs are probabilistic analyses that estimate the core damage frequency (CDF) and containment performance for accidents initiated by internal events. The US Nuclear Regulatory Commission (NRC) reviewed the IPE submittals with the objective of gaining perspectives in three major areas: (1) improvements made to individual plants as a result of their IPEs and the collective results of the IPE program, (2) plant-specific design and operational features and modeling assumptions that significantly affect the estimates of CDF and containment performance, and (3) strengths and weaknesses of the models and methods used in the IPEs. These perspectives are gained by assessing the core damage and containment performance results, including overall CDF, accident sequences, dominant contributions to component failure and human error, and containment failure modes. Methods, data, boundary conditions, and assumptions used in the IPEs are considered in understanding the differences and similarities observed among the various types of plants. This report is divided into three volumes containing six parts. Part 1 is a summary report of the key perspectives gained in each of the areas identified above, with a discussion of the NRC`s overall conclusions and observations. Part 2 discusses key perspectives regarding the impact of the IPE Program on reactor safety. Part 3 discusses perspectives gained from the IPE results regarding CDF, containment performance, and human actions. Part 4 discusses perspectives regarding the IPE models and methods. Part 5 discusses additional IPE perspectives. Part 6 contains Appendices A, B and C which provide the references of the information from the IPEs, updated PRA results, and public comments on draft NUREG-1560 respectively.

  14. Human Reliability Analysis for Small Modular Reactors

    SciTech Connect (OSTI)

    Ronald L. Boring; David I. Gertman

    2012-06-01T23:59:59.000Z

    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.

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

    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)

    1994-08-01T23:59:59.000Z

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

  16. 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)

    1995-03-01T23:59:59.000Z

    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.

  17. System Effectiveness

    SciTech Connect (OSTI)

    Powell, Danny H [ORNL] [ORNL; Elwood Jr, Robert H [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    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.

  18. 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

    2010-02-22T23:59:59.000Z

    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.

  19. Identification and Assessment of Material Models for Age-Related Degradation of Structures and Passive Components in Nuclear Power Plants

    SciTech Connect (OSTI)

    Nie,J.; Braverman, J.; Hofmayer, C.; Kim, M. K.; Choi, I-K.

    2009-04-27T23:59:59.000Z

    When performing seismic safety assessments of nuclear power plants (NPPs), the potential effects of age-related degradation on structures, systems, and components (SSCs) should be considered. To address the issue of aging degradation, the Korea Atomic Energy Research Institute (KAERI) has embarked on a five-year research project to develop a realistic seismic risk evaluation system which will include the consideration of aging of structures and components in NPPs. Three specific areas that are included in the KAERI research project, related to seismic probabilistic risk assessment (PRA), are probabilistic seismic hazard analysis, seismic fragility analysis including the effects of aging, and a plant seismic risk analysis. To support the development of seismic capability evaluation technology for degraded structures and components, KAERI entered into a collaboration agreement with Brookhaven National Laboratory (BNL) in 2007. The collaborative research effort is intended to continue over a five year period with the goal of developing seismic fragility analysis methods that consider the potential effects of age-related degradation of SSCs, and using these results 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 that will be performed in the subsequent evaluations in the years that follow. 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. This report describes the research effort performed by BNL for the Year 2 scope of work. This research focused on methods that could be used to represent the long-term behavior of materials used at NPPs. To achieve this BNL reviewed time-dependent models which can approximate the degradation effects of the key materials used in the construction of structures and passive components determined to be of interest in the Year 1 effort. The intent was to review the degradation models that would cover the most common time-dependent changes in material properties for concrete and steel components.

  20. Summary

    SciTech Connect (OSTI)

    Powell, Danny H [ORNL] [ORNL; Elwood Jr, Robert H [ORNL] [ORNL

    2011-01-01T23:59:59.000Z

    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. The material control and accountability (MC&A) system effectiveness tool (MSET) is a self-assessment or inspection tool utilizing probabilistic risk assessment (PRA) methodology to calculate the system effectiveness of a nuclear facility's material protection, control, and accountability (MPC&A) system. The MSET process is divided into four distinct and separate parts: (1) Completion of the questionnaire that assembles information about the operations of every aspect of the MPC&A system; (2) Conversion of questionnaire data into numeric values associated with risk; (3) Analysis of the numeric data utilizing the MPC&A fault tree and the SAPHIRE computer software; and (4) Self-assessment using the MSET reports to perform the effectiveness evaluation of the facility's 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. If the need for system improvements or upgrades is indicated when the system is analyzed, MSET provides the capability to evaluate potential or actual system improvements or upgrades. A facility's MC&A system can be evaluated at a point in time. The system can be 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 system improvement 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 and reveals where performance degradation has the greatest impact on total system risk. The risk importance factors show the amount of risk reduction achievable with potential upgrades and the amount of risk reduction 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. The functional model, the system risk assessment tool, and the facility evaluation questionnaire are valuable educational tools for MPC&A personnel. These educational tools provide a framework for ongoing dialogue between organizations regarding the design, development, implementation, operation, assessment, and sustainability of MPC&A systems. An organization considering the use of MSET as an analytical tool for evaluating the effectiveness of its MPC&A system will benefit from conducting a complete MSET exercise at an existing nuclear facility.

  1. 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.

    2010-06-30T23:59:59.000Z

    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.