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. 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

  3. 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.

  4. 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.

  5. 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.

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

    SciTech Connect (OSTI)

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

    1990-01-01T23:59:59.000Z

    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.

  7. 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.

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

    Energy Savers [EERE]

    an enduring data and modeling resource to minimize duplication of effort across DOE and train future generation of PA professionals In late 2013, the group was broadened as P&RA...

  9. PRA Anoplophora chinensis, Plant Protection Service, Wageningen, The Netherlands, September 2008 1 Pest Risk Analysis

    E-Print Network [OSTI]

    PRA Anoplophora chinensis, Plant Protection Service, Wageningen, The Netherlands, September 2008 1, Wageningen, The Netherlands, September 2008 2 European and Mediterranean Plant Protection Organisation Franck Hérard Plant Protection Service, P.O. Box 9102, 6700 HC Wageningen, The Netherlands Plant

  10. 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.

  11. 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.

  12. 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.

  13. 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.

  14. 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.

  15. DOE safety goals comparison using NUREG-1150 PRA (probabilistic risk assessment) methodology

    SciTech Connect (OSTI)

    Wang., O.S.; Zentner, M.D.; Rainey, T.E.

    1990-06-01T23:59:59.000Z

    A full-scope Level 3 probabilistic risk assessment (PRA) including external events has been performed for N Reactor, a US Department of Energy (DOE) Category A production reactor. This four-year, multi-million dollar task was a joint effort by Westinghouse Hanford Company, Science Applications International Corporation (SAIC), and Sandia National Laboratories (SNL). Technical lead in external events and NUREG-1150 methodology was provided by SNL. SAIC led the effort in the Level 1 analysis for the internally initiated events. Westinghouse Hanford supported the task in many key areas, such as data collection and interpretation, accident progression, system interaction, human factor analyses, expert elicitation, peers review, etc. The main objective of this Level 3 PRA are to assess the risks to the public and onsite workers posed by the operation of N Reactor, to identify modifications to the plant that could reduce the overall risk, and to compare those risks to the proposed DOE and Nuclear Regulatory Commission (NRC) quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford and SNL for estimating individual health risks, and the comparison of the N Reactor results and DOE quantitative nuclear safety guidelines. This paper is devoted to DOE quantitative safety guidelines interpretation and comparison; the NRC safety objectives are also presented in order to compare N Reactor results to commercial nuclear power plants included in the NUREG-1150 study. 7 refs., 7 tabs.

  16. 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

  17. 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.

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

    E-Print Network [OSTI]

    Norris, Ray

    PoS(PRA2009)028 The ATLAS Survey of the CDFS and ELAIS-S1 Fields Emil Lenc, Ray Norris Australia Telescope National Facility E-mail: Emil.Lenc@csiro.au, Ray.Norris@csiro.au Andrew Hopkins, Rob Sharp Anglo of Sydney E-mail: krandall@physics.usyd.edu.au The first phase of the ATLAS (Australia Telescope Large Area

  19. The use of PRA (Probabilistic Risk Assessment) in the management of safety issues at the High Flux Isotope Reactor

    SciTech Connect (OSTI)

    Flanagan, G.F.

    1990-01-01T23:59:59.000Z

    The High Flux Isotope reactor (HFIR) is a high performance isotope production and research reactor which has been in operation at Oak Ridge National Laboratory (ORNL) since 1965. In late 1986 the reactor was shut down as a result of discovery of unexpected neutron embrittlement of the reactor vessel. In January of 1988, a level 1 Probabilistic Risk Assessment (PRA) (excluding external events) was published as part of the response to the many reviews that followed the shutdown and for use by ORNL to prioritize action items intended to upgrade the safety of the reactor. A conservative estimate of the core damage frequency initiated by internal events for HFIR was 3.11 {times} 10{sup {minus}4}. In June 1989 a draft external events initiated PRA was published. The dominant contributions from external events came from seismic, wind, and fires. The overall external event contribution to core damage frequency is about 138% of the internal event initiated contribution and is dominated by wind initiators. The PRA has provided a basis for the management of a wide range of safety and operation issues at the HFIR. 3 refs., 4 figs., 2 tabs.

  20. 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.

  1. 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.

  2. 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...

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

    E-Print Network [OSTI]

    of controlled thermonuclear fusion research [1, 2, 3, 4, 5]. These structures deserve a special interest

  4. Technology acquisition: sourcing technology from industry partners

    E-Print Network [OSTI]

    Ortiz-Gallardo, Victor Gerardo

    2013-07-09T23:59:59.000Z

    gaps - Pra ctice- based Framew ork (v1.0). - 13 S emi- structu red intervie ws / Ground ed anal ysis CHAPTER  1 9 Cha pterP urpose of the chapte r Main a rgume nts/fin dings Frame work version Data s ources / Analyt ical me thod... ses sion where the ref ined fr amewo rk (v3. 0) was presen ted to a fo rum of practi tioners . - The c omme nts and feedb ack pr ovided by par ticipan ts supp orted t he results of the resear ch. - - - Foc us gro up. - Par ticipan ts...

  5. 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

  6. 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...

  7. 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...

  8. 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

  9. 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.

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

    Office of Environmental Management (EM)

    45 Roger Seitz SRNL 46 David Sevougian SNL Name AgencyCompany Affliation 47 Greg Shott NSTec 48 Linda Suttora DOE EM 49 John Tauxe Neptune 50 Candice Trummell DOE EM 51 David Ward...

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

    Office of Environmental Management (EM)

    Waste Landfill Performance Assessment Updates for New Waste Streams, Mr. Greg Shott (NSTec) 3:15 - 3:30 pm Break 3:30 - 4:30 pm Use of Probabilistic Performance Assessment in...

  12. 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.

  13. An evaluation of internal event level 1 PRA methods used in NUREG-1150

    SciTech Connect (OSTI)

    Camp, A.L.; Cramond, W.R.

    1989-01-01T23:59:59.000Z

    As part of the effort to support NUREG-1150, Sandia National Laboratories and its subcontractors have developed innovative techniques for efficiently performing internal event Level I probabilistic risk assessments. This methodology is one of the alternatives for industry to use in performing individual plant evaluations in the future. While this new methodology was very successful, there are some areas where improvements can be made. This paper evaluates the strengths and weaknesses of the methodology and makes some important recommendations for modifications in order to provide insights to future users. 10 refs.

  14. Microsoft Word - 2011-09-16 - PRA Paper - Final Clean - Rev 1...

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

    1270 Next Generation Nuclear Plant Probabilistic Risk Assessment White Paper September 2011 DISCLAIMER This information was prepared as an account of work sponsored by an agency of...

  15. PoS(PRA2009)065 Galaxy transformation in dense environments: A

    E-Print Network [OSTI]

    Kraan-Korteweg, Renée C.

    from the low-density field population of galaxies in the local Universe which are predominantly spirals der Heyden Astronomy Department, Univ. of Cape Town, Private Bag X3, Rondebosch 7701, South Africa E, Observatory 7935, South Africa U. Fritze, R. Kotulla Centre for Astrophysics Research, Univ. of Hertfordshire

  16. Level 3 PRA Reoprt.DRAFT.9.13.14.0030.docx.docx

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

    dispersion model tend to over predict radionuclide concentrations, and under predict radionuclide spread. 6 The Hybrid Single Particle Lagrangian Integrated Trajectory...

  17. 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...

  18. ht. 1. Han Mass 7h&r. Vol. 13, pp. 13494357. Pergamon Pra 1970. PhIed in Great Britain RADIATIVE TRANSFER IN A CONSERVATIVE

    E-Print Network [OSTI]

    Siewert, Charles E.

    involving radiative transport and wall temperature slip in a finite, absorbing, emitting gray medium, equation of transfer then equation (4) reduces to the simpler form where Z(z,Zl)is the radiation intensity, Zlis the In their work on radiative transport and wall direction cosine (as measured from the positive

  19. 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

  20. 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

  1. 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...

  2. 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

  3. REVUE DE PHYSIQUE APPLIQUE The least square method in the determination

    E-Print Network [OSTI]

    Boyer, Edmond

    83 REVUE DE PHYSIQUE APPLIQUÃ?E The least square method in the determination of thermal diffusivity al. [1] used in thermal diffusivity measurements by a laser flash method. The presented method of the most popular to measure thermal diff'usivity of solids. The oldest and simplest model on which

  4. Transport Analysis of Radial Electric Field in Helical Plasmas

    E-Print Network [OSTI]

    condition for the neoclassical particle ux. The generation of the electric #12;eld in helical systems could in generating the radial electric #12;eld [8, 9]. We have used the transport model for anomalous di#11;usivitiesTransport Analysis of Radial Electric Field in Helical Plasmas S. Toda and K. Itoh National

  5. 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

  6. HYDRODYNAMIC LIMITS FOR KINETIC EQUATIONS AND THE DIFFUSIVE APPROXIMATION OF RADIATIVE

    E-Print Network [OSTI]

    Tzavaras, Athanasios E.

    HYDRODYNAMIC LIMITS FOR KINETIC EQUATIONS AND THE DIFFUSIVE APPROXIMATION OF RADIATIVE TRANSPORT . The radiative transport equations, satisfied by the Wigner function for random acoustic waves, present#usive approximation of the radiative transport equation. 1. Introduction We consider a class of kinetic models

  7. 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

  8. 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

  9. 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

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

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

    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...

  11. Infrastructure Security EXCEPTIONAL SERVICE IN THE NATIONAL INTEREST

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

    PRA research and analysis expertise including national infrastructure, transportation, health care, agriculture, and, of course, nuclear power and national defense....

  12. 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)  

  13. 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...

  14. 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...

  15. 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.

  16. External Technical Reviews | 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...

  17. E-Print Network 3.0 - aldosterone renin activity Sample Search...

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

    Angiotensin II and Aldosterone Increase with Fasting in Breeding Adult Male Northern Elephant Seals (Mirounga angustirostris) Summary: renin activity (PRA), an indicator of...

  18. Information Collection Management | Department of Energy

    Office of Environmental Management (EM)

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

  19. 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...

  20. 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...

  1. Interstate Technology & Regulatory Council (ITRC) Remediation...

    Office of Environmental Management (EM)

    Laboratory December 17, 2014 To view all the P&RA CoP 2014 Technical Exchange Meeting videos click here. Video Presentation Interstate Technology & Regulatory Council (ITRC)...

  2. 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

  3. E-Print Network 3.0 - accident sequence analysis Sample Search...

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

    , Preliminary risk analysis (PRA), risk, potential accident, feared events, Automatic Train Control. I... , dangers and potential accidents respectively. At the beginning of the...

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

    Office of Environmental Management (EM)

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

  5. 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...

  6. Proceedings of the US Nuclear Regulatory Commission fourteenth water reactor safety information meeting: Volume 1, Plenary session, Severe accident sequence analysis, Risk analysis/PRA applications, Reference plant risk analysis - NUREG-1150, Innovative concepts for increased safety of advanced power reactors

    SciTech Connect (OSTI)

    Weiss, A.J. (comp.)

    1987-02-01T23:59:59.000Z

    This six-volume report contains 156 papers out of the 175 that were presented at the Fourteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 27-31, 1986. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. Foreign participation in the meeting included thirty-four different papers presented by researchers from Canada, Czechoslovakia, Finland, Germany, Italy, Japan, Mexico, Spain, Sweden, Switzerland and the United Kingdom. The titles of the papers and the names of the authors have been updated and may differ from those that appeared in the final program of the meeting.

  7. Proceedings of the US Nuclear Regulatory Commission fifteenth water reactor safety information meeting: Volume 1, Plenary sessions, reactor licensing topics, NUREG-1150, risk analysis/PRA applications, innovative concepts for increased safety of advanced power reactors, severe accident modeling and analysis

    SciTech Connect (OSTI)

    Weiss, A.J. (comp.)

    1988-02-01T23:59:59.000Z

    This six-volume report contains 140 papers out of the 164 that were presented at the Fifteenth Water Reactor Safety Information Meeting held at the National Bureau of Standards, Gaithersburg, Maryland, during the week of October 26-29, 1987. The papers are printed in the order of their presentation in each session and describe progress and results of programs in nuclear safety research conducted in this country and abroad. This report, Volume 1, discusses the following: plenary sessions; reactor licensing; NUREG-1150; risk analysis; innovative concepts for increased safety of advanced power reactors; and severe accident modeling and analysis. Thirty-two reports have been cataloged separately.

  8. 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

  9. The genome and cytoskeleton of Naegleria gruberi, an amoeboflagellate

    E-Print Network [OSTI]

    Fritz-Laylin, Lillian Kathleen

    2010-01-01T23:59:59.000Z

    Neurospora crassa, hsa human, tad Trichoplax adherens, mbrID) in family FM1 pra,hsa,ppa,mbr,tad,tps,p te,tbr,gla,tva,cre,ngr FM2 pra,hsa,ppa,mbr,tad,tps,p te,tbr,gla,tva,cre,ngr

  10. 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

  11. 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.

  12. 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 ...

  13. Sabotaging Logics: How Brazil's Hip-Hop Culture Looks to Redefine Race

    E-Print Network [OSTI]

    Moulin, Maria Teresa

    2010-01-01T23:59:59.000Z

    to history’s importance: Vim pelo caminho difícil, a linhaasunto aqui é o crime, eu vim aqui por isso…” ‘But the issuepacífico, verídico, vim pra sabotar seu raciocínio E a

  14. 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...

  15. 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. ...

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

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

    IROFS are viewed as being of equal importance. On the other hand, a PRA can be used to rank IROFS in terms of risk importance measures. Without such rankings, licensees and...

  17. 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...

  18. E-Print Network 3.0 - australia telescope facilities Sample Search...

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

    Large Area... -NonCommercial-ShareAlike Licence. http:pos.sissa.it 12;PoS(PRA2009)028 ATLAS ... Source: Norris, Ray - Australia Telescope National Facility, CSIRO Collection:...

  19. 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...

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

    Office of Environmental Management (EM)

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

  1. 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...

  2. 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...

  3. 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...

  4. 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...

  5. 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...

  6. 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...

  7. 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....

  8. 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 ...

  9. February 5, 2014 Webinar - The Cementitious Barriers Partnership...

    Office of Environmental Management (EM)

    of the Cementitious Barriers Partnership Toolbox, Version 2.0 David Kosson et al. (Vanderbilt UniversityCRESP) Agenda - 252014 P&RA CoP Webinar Presentation - Tools...

  10. 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 ...

  11. 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 ...

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

    SciTech Connect (OSTI)

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

    1991-11-01T23:59:59.000Z

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

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

    SciTech Connect (OSTI)

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

    1991-11-01T23:59:59.000Z

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

  14. 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...

  15. 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.

  16. 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.

  17. 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.

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

    SciTech Connect (OSTI)

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

    1988-01-01T23:59:59.000Z

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

  19. 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.

  20. 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.

  1. The Zion integrated safety analysis for NUREG-1150

    SciTech Connect (OSTI)

    Unwin, S.D.; Park, C.K.

    1988-01-01T23:59:59.000Z

    The utility-funded Zion Probabilistic Safety Study provided not only a detailed and thorough assessment of the risk profile of Zion Unit 1, but also presented substantial advancement in the technology of probabilistic risk assessment (PRA). Since performance of that study, modifications of plant hardware, the introduction of new emergency procedures, operational experience gained, information generated by severe accident research programs and further evolution of PRA and uncertainty analysis methods have provided a basis for reevaluation of the Zion risk profile. This reevaluation is discussed in this report. 5 refs.

  2. 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

  3. Development of a Modelling and Simulation Method Comparison and Selection Framework for Health Services Management

    E-Print Network [OSTI]

    Jun, Gyuchan T; Morris, Zoe; Eldabi, Tillal; Harper, Paul; Naseer, Aisha; Patel, Brijesh; Clarkson, John P

    2011-05-19T23:59:59.000Z

    , Georgia, USA Edited by: Andradottir S, Healy KJ, Withers DH, Nelson BL 1997. 13. Singh AJ, May PRA: Acceptance of operations research/systems analysis in the health care field. Interfaces 1977, 7:79-86. 14. Linder JA, Rose AF, Palchuk MB, Chang F...

  4. 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

  5. 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.

  6. A Complete Axiomatization of Differential Game Logic for Hybrid Games

    E-Print Network [OSTI]

    is an updated version superseding the earlier report CMU-CS-12-105 [Pla12b]. It is based on that earlier report logic (dL) [Pla08, Pla12a], which extends Pratt's dynamic logic of conventional discrete programs [Pra76 introduce differential game logic (dGL) [Pla12b] for studying the existence of winning strategies for hybrid

  7. 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

  8. 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...

  9. 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

  10. 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

  11. The School of Social Sciences Degree Offered: BA

    E-Print Network [OSTI]

    Richards-Kortum, Rebecca

    observation; and case studies of a variety of social phenomena, processes, and problems as methods Steve H. Murdock Professors emeriti Chandler Davidson William Martin Professor in the PraCtiCe a sustained independent research project. Eligibility--To be eligible for the program, students must have

  12. 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

  13. Commissioner George Apostolakis U.S. Nuclear Regulatory Commission

    E-Print Network [OSTI]

    Bernstein, Joseph B.

    (Reactors) · Study the system as an integrated socio- technical system · Probabilistic Risk Assessment (PRA Enrichment Power Reactors Transportation Storage Waste Disposal Uranium Conversion Medical/Industrial #12;The Traditional Approach to Regulation (Before Risk Assessment) · Management of uncertainty (unquantified

  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. Probabilistic risk assessment of N Reactor using NUREG-1150 methods

    SciTech Connect (OSTI)

    Wang, O.S.; Baxter, J.T.; Coles, G.A.; Powers, T.B.; Zentner, M.D.

    1989-11-01T23:59:59.000Z

    A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a US Department of Energy (DOE) Category A production reactor. The main contractor is Westinghouse Hanford Company (Westinghouse Hanford). The PRA methodology developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL) in support of the NUREG-1150 (Reference 1) effort were used for this analysis. N Reactor is a graphite-moderated pressurized water reactor designed by General Electric. The dual-purpose 4000 MWt nuclear plant is located within the Hanford Site in the south-central part of the State of Washington. In addition to producing special materials for the DOE, N Reactor generates 860 MWe for the Washington Public Power Supply System. The reactor has been operated successfully and safely since 1963, and was put into standby status in 1988 due to the changing need in special nuclear material. 3 refs., 4 tabs.

  20. Expert opinion in risk analysis: The NUREG-1150 methodology

    SciTech Connect (OSTI)

    Hora, S.C.; Iman, R.L.

    1988-01-01T23:59:59.000Z

    The Reactor Risk Reference Document (US Nuclear Regulatory Commission, 1987) is the most comprehensive study and application of probabilistic risk analysis and uncertainty analysis methods for nuclear power generation safety since the Reactor Safety Study (US Nuclear Regulatory Commission, 1975). Many of the issues addressed in PRA work such as NUREG-1150 involve phenomena that have not been studied through experiment or observation to an extent that makes possible a definitive analysis. In many instances, the rarity or severity of the phenomena make resolution impossible at this time. In these instances, the best available information resides with experts who have studied the phenomena in question. This paper is about a reasoned approach to the acquisition of expert opinion for use in PRA work and other public policy areas.

  1. Cable Hot Shorts and Circuit Analysis in Fire Risk Assessment

    SciTech Connect (OSTI)

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

    1999-05-19T23:59:59.000Z

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

  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. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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)

  8. Uncertainty associated with probabilistic prediction of nutrient transport by runoff

    E-Print Network [OSTI]

    Jain, Mohit

    1996-01-01T23:59:59.000Z

    and estimating the probability and severity of potential hazards to water quality. The objective of this research was to use PRA to characterize the uncertainty associated with probabilistic determination of the nutrient transport by runoff at two dairies.... Simple simulation models were used to determine the rainfall runoff probability and lagoon overflow probabilities. Phosphorous index method in combination with nutrient application rates and soil test levels was used to determine the presence oF excess...

  9. Generalized qudit Choi maps

    E-Print Network [OSTI]

    M. A. Jafarizadeh; M. Rezaeen; S. Ahadpour

    2006-07-24T23:59:59.000Z

    Following the linear programming prescription of Ref. \\cite{PRA72}, the $d\\otimes d$ Bell diagonal entanglement witnesses are provided. By using Jamiolkowski isomorphism, it is shown that the corresponding positive maps are the generalized qudit Choi maps. Also by manipulating particular $d\\otimes d$ Bell diagonal separable states and constructing corresponding bound entangled states, it is shown that thus obtained $d\\otimes d$ BDEW's (consequently qudit Choi maps) are non-decomposable in certain range of their parameters.

  10. Risk management at the Oak Ridge National Laboratory research reactors

    SciTech Connect (OSTI)

    Flanagan, G.F.; Linn, M.A.; Proctor, L.D.; Cook, D.H.

    1994-12-31T23:59:59.000Z

    In November of 1986, the High Flux Isotope Reactor (HFIR) was shut down by Oak Ridge National Laboratory (ORNL) due to a concern regarding embrittlement of the reactor vessel. A massive review effort was undertaken by ORNL and the Department of Energy (DOE). This review resulted in an extensive list of analyses and design modifications to be completed before restart could take place. The review also focused on the improvement of management practices including implementation of several of the Institute of Nuclear Power Operations (INPO) requirements. One of the early items identified was the need to perform a Probabilistic Risk Assessment (PRA) on the reactor. It was decided by ORNL management that this PRA would not be just an exercise to assess the ``bottom`` line in order to restart, but would be used to improve the overall safety of the reactor, especially since resources (both manpower and dollars) were severely limited. The PRA would become a basic safety tool to be used instead of a more standard deterministic approach to safety used in commercial reactor power plants. This approach was further reinforced, because the reactor was nearly 25 years old at this time, and the design standards and regulations had changed significantly since the original design, and many of the safety issues could not be addressed by compliance to codes and standards.

  11. Advanced neutron source reactor probabilistic flow blockage assessment

    SciTech Connect (OSTI)

    Ramsey, C.T.

    1995-08-01T23:59:59.000Z

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

  12. 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.

  13. 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.

  14. N Reactor external events probabilistic risk assessment using NUREG-1150 methods

    SciTech Connect (OSTI)

    Wang, O.S.; Baxter, J.T.; Coles, G.A.; Zentner, M.D.; Powers, T.B.; Collard, L.B.; Rainey, T.E.

    1990-01-01T23:59:59.000Z

    This is the first full-scope Level-III PRA completed for the DOE Category A reactor using the updated NUREG-1150 methods. The comparisons to the quantitative NRC safety objectives and DOE nuclear safety guidelines also set analytical precedent for DOE production reactors. Generally speaking, the risks of operating N Reactor are low because of a combination of factors such as low power density, large confinement volume, effective redundant scram systems and core cooling systems, remote location, etc. This work has been a major effort to evaluate the N Reactor risk using state-of-the-art PRA technology. It is believed that this PRA has resulted in realistic, or slightly conservative, results (as opposed to unduly conservative or nonconservative results). The study concluded that the risk to the public and to nearby DOE workers from the operation of N Reactor is very low. This analysis also found that N Reactor meets all the quantitative NRC safety objectives and DOE nuclear safety guidelines, and is generally as safe as, or safer than most commercial reactors in terms of societal and individual risks. The calculated risk to Hanford onsite workers is comparable to public risk from commercial reactors in the NUREG-1150 study. As a result of these low-risk estimates, only a small effort has been devoted to identifying significant risk reduction alternatives. 22 refs., 2 figs., 10 tabs.

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

    SciTech Connect (OSTI)

    Moffitt, N.E.; Gore, B.F.: Vo, T.V. (Pacific Northwest Lab., Richland, WA (USA))

    1991-07-01T23:59:59.000Z

    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. Byron and Braidwood were selected for the fourth study in this program. The produce of this effort is a prioritized listing of AFW failures which have occurred at the plants and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the Byron/Braidwood plants. 23 refs., 1 fig., 1 tab.

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

    SciTech Connect (OSTI)

    Moffitt, N.E.; Lloyd, R.C.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States); Garner, L.W. [Nuclear Regulatory Commission, Washington, DC (United States)

    1993-08-01T23:59:59.000Z

    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. H. B. Robinson was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the H. B. Robinson plant.

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

    SciTech Connect (OSTI)

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

    1994-05-01T23:59:59.000Z

    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. McGuire was selected as one of a series of plants for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important components at the McGuire plant.

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

    SciTech Connect (OSTI)

    Vo, T.V.; Pugh, R.; Gore, B.F.; Harrison, D.G. (Pacific Northwest Lab., Richland, WA (USA))

    1990-10-01T23:59:59.000Z

    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 recently developed for the NRC for various plants was used to identify generic component failure modes. This information was then combined with plant-specific and industry-wide component information and failure data to identify failure modes and failure mechanisms for the AFW system at the selected plants. J. M. Farley was selected as the second plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by NRC inspectors in the preparation of inspection plans addressing AFW risk-important at the J. M. Farley plant. 23 refs., 1 fig., 1 tab.

  19. 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.

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

    SciTech Connect (OSTI)

    Bumgardner, J.D.; Nickolaus, J.R.; Moffitt, N.E.; Gore, B.F.; Vo, T.V. [Pacific Northwest Lab., Richland, WA (United States)

    1993-12-01T23:59:59.000Z

    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. South Texas Project was selected as a plant for study. The product of this effort is a prioritized listing of AFW failures which have occurred at the plant and at other PWRs. This listing is intended for use by the NRC inspectors in preparation of inspection plans addressing AFW risk important components at the South Texas Project plant.

  1. 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.

  2. Level III probabilistic risk assessment for N Reactor

    SciTech Connect (OSTI)

    Camp, A.L.; Kunsman, D.M.; Miller, L.A.; Sprung, J.L.; Wheeler, T.A.; Wyss, G.D. (Sandia National Labs., Albuquerque, NM (USA))

    1990-04-01T23:59:59.000Z

    A Level III probabilistic risk assessment (PRA) has been performed for N Reactor, a Department of Energy (DOE) production reactor located on the Hanford reservation in Washington. The objectives of the PRA are to assess the risks to the public and the Hanford site workers posed by the operation of N Reactor, to compare those risks to proposed DOE safety goals, and to identify changes to the plant that could reduce the risk. The scope of the PRA is comprehensive, excluding only sabotage and operation errors of commission. State-of-the-art methodology is employed based largely on the methods developed by Sandia for the US Nuclear Regulatory Commission in support of the NUREG-1150 study of five commercial nuclear power plants. The structure of the probabilistic models allowed complex interactions and dependencies between systems to be explicitly considered. Latin Hypercube sampling techniques were used to develop uncertainty distributions for the risks associated with postulated core damage events initiated by fire, seismic, and internal events as well as the overall combined risk. The combined risk results show that N Reactor meets the primary DOE safety goals and compared favorably to the plants considered in the NUREG-1150 analysis. 36 figs., 81 tabs.

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

    SciTech Connect (OSTI)

    Wyss, G.D.; Camp, A.L.; Miller, L.A.; Dingman, S.E.; Kunsman, D.M. (Sandia National Labs., Albuquerque, NM (USA)); Medford, G.T. (Science Applications International Corp., Albuquerque, NM (USA))

    1990-06-01T23:59:59.000Z

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

  4. 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.

  5. Systems Analysis Programs for Hands-on Intergrated Reliability Evaluations (SAPHIRE) Summary Manual

    SciTech Connect (OSTI)

    C. L. Smith

    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 (PC) running the Microsoft Windows operating system. SAPHIRE is primarily funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). INL's primary role in this project is that of software developer and tester. However, INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users, who constitute a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events and quantify associated consequential outcome frequencies. Specifically, for nuclear power plant applications, SAPHIRE can identify important contributors to core damage (Level 1 PRA) and containment failure during a severe accident which lead to releases (Level 2 PRA). It can be used for a PRA where the reactor is at full power, low power, or at shutdown conditions. Furthermore, it can be used to analyze both internal and external initiating events and has special features for transforming an internal events model to a model for external events, such as flooding and fire analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to the public and environment (Level 3 PRA). SAPHIRE also includes a separate module called the Graphical Evaluation Module (GEM). GEM is a special user interface linked to SAPHIRE that automates the SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events (for example, to calculate a conditional core damage probability) very efficiently and expeditiously. This report provides an overview of the functions available in SAPHIRE and presents general instructions for using the software. Section 1 presents SAPHIRE’s historical evolution and summarizes its capabilities. Section 2 presents instructions for installing and using the code. Section 3 explains the database structure used in SAPHIRE and discusses database concepts. Section 4 explains how PRA data (event frequencies, human error probabilities, etc.) can be generated and manipulated using “change sets.” Section 5 deals with fault tree operations, including constructing, editing, solving, and displaying results. Section 6 presents operations associated with event trees, including rule application for event tree linking, partitioning, and editing sequences. Section 7 presents how accident sequences are generated, solved, quantified, and analyzed. Section 8 discusses the functions available for performing end state analysis. Section 9 explains how to modify data stored in a SAPHIRE database. Section 10 illustrates how to generate and customize reports. Section 11 covers SAPHIRE utility options to perform routine functions such as defining constant values, recovering databases, and loading data from external sources. Section 12 provides an overview of GEM’s features and capabilities. Finally, Section 13 summarizes SAPHIRE’s quality assurance process.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. Fractionation studies on the unidentified growth factor(s) in distillers dried solubles

    E-Print Network [OSTI]

    Dannenburg, Warren Nathaniel

    1955-01-01T23:59:59.000Z

    . This was called "methyl aloohol soluble fraction of distillers dried solublesi The residue wss air drie4 and labeled "aetna 1 alcohol insoluble fraotion of distillers drie4 solubles". $. r fo m at nt ist e i o ubl s Five hundred gm of distillexs dried... fraction ox Ms- tillers Cried solubles (pH 1)"L "water soluole fr~ction of distillers dried solubles (PH '/)"L ~ "water soluble fxaction of dis~illers dried solubles (pH 11)". ur h r Pra tionatio f th Sate 8 lub e }raut of 9 still rs ed Soluo es a...

  12. 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.

  13. 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.

  14. The GTE Ceramic Recuperator for High Temperature Waste Heat Recovery

    E-Print Network [OSTI]

    Dorazio, R. E.; Gonzalez, J. M.; Ferri, J. L.; Rebello, W. J.; Ally, M. R.

    1984-01-01T23:59:59.000Z

    Steel Bllffalo Metal Casting Standard St.eel N.ati_onal Forge Ladish Co. Pr.Jt.t & \\.fllitney Ama", Specl."11t.v Metals Bethlehem Steel Cape Ann Forge Staolev Spring (TRw) Box Forge Reheat, Steel Box Forge Reheat, Steel 1 Box Forge Reheat...,807 1.9 1.8 31 St.andard Steel Burnham, PA Box forge. Reheat, Steel 32 National Forge Erie, PA Ladle Preheater. Steel :,.} Lad isb Co. Cyntbiaca, ....'Y Box Heat Treat, Steell 188.426 77,527 3. Pra t t & \\.on i tney East Hart.ford, CT Box...

  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. 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.

  17. Calcium sensitivity determinations by neutron activation analysis as applied to bone

    E-Print Network [OSTI]

    Blasdel, Michael John

    2012-06-07T23:59:59.000Z

    . ated, eliminating a total body dose. But primarily, the ettuipment used is greatly reduced in size and cost froa~ praY. . ou" roric?u. in' a poa tai~lc neutron aource (againat t', u uacs of a cyclo' ton) aad on1 y onu acinti llatf on cry tnl...PII(IILiL'1 of r il 'lto. l Iir, 'L 'v;il'(i(i', N ~ ? - (1-e ) &5 Nc& i a e(5uation 1 N ? null&'~er of radioactive atoms present at end of irradiation (atoms) - neutron flux (neutrons/cm 'sec) 2. N . - total nuAer of orig" nal atoms (atoms) 1 ? decay...

  18. Sundry Brief Articles, Compiled from Press Notes Published During the Years 1894 and 1895. Index.

    E-Print Network [OSTI]

    Connell, J. H.; Harrington, H. H.; Francsis, M.; Price, R. H.; Clayton, James

    1895-01-01T23:59:59.000Z

    little), and 662 TEXAS AGRICULTURAL EXPERIMENT STATION. that Lh butter g rained nicely. But J have also found out that the feed in()' of cotLon d meal and hulls free ly and continuou ly will ruin a dairy herd. The cow under uch conditions fail... and to the height of the mercury. These facts cou ll bave come to li o?ht lo ng since were it not for the fact that the ext n ive u e of cotton seed and cotton seed meal for feeding purposes is a n w pra tice. If fed in moderation my experience teaches me...

  19. Commercial Fertilizers in 1933-34.

    E-Print Network [OSTI]

    Fraps, G. S. (George Stronach); Asbury, S. E. (Samuel E.)

    1934-01-01T23:59:59.000Z

    . S., Entomologist H. B. Parks, B. S., Chief **E. W. Dunnam, Ph. D., Entomologist A. H. Alex, B. S., Queen Breeder **R. W. Moreland. B. S., Asst. Entomologist Feed Control Service: C. E. Heard, B. S., Chief Inspector F. D. Fuller, M. S., Chief C. J..., 192 **In cooperation with U. S. Department of Agriculture. $In cooperation with Texas Extension Service. YNOPSI! '1 tics ferl -diff f aci 33,4 as Pra 20 i T and 'h~s is the annual Fertilizer Control Bulletin. It contains stal...

  20. 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...

  1. 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...

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

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

    SciTech Connect (OSTI)

    Flanagan, G.F.; Johnson, D.H.; Buttemer, D.; Perla, H.F.; Chien, S.H. (Oak Ridge National Lab., TN (USA))

    1989-01-01T23:59:59.000Z

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

  8. 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.

  9. Multi-State Physics Models of Aging Passive Components in Probabilistic Risk Assessment

    SciTech Connect (OSTI)

    Unwin, Stephen D.; Lowry, Peter P.; Layton, Robert F.; Heasler, Patrick G.; Toloczko, Mychailo B.

    2011-03-13T23:59:59.000Z

    Multi-state Markov modeling has proved to be a promising approach to estimating the reliability of passive components - particularly metallic pipe components - in the context of probabilistic risk assessment (PRA). These models consider the progressive degradation of a component through a series of observable discrete states, such as detectable flaw, leak and rupture. Service data then generally provides the basis for estimating the state transition rates. Research in materials science is producing a growing understanding of the physical phenomena that govern the aging degradation of passive pipe components. As a result, there is an emerging opportunity to incorporate these insights into PRA. This paper describes research conducted under the Risk-Informed Safety Margin Characterization Pathway of the Department of Energy’s Light Water Reactor Sustainability Program. A state transition model is described that addresses aging behavior associated with stress corrosion cracking in ASME Class 1 dissimilar metal welds – a component type relevant to LOCA analysis. The state transition rate estimates are based on physics models of weld degradation rather than service data. The resultant model is found to be non-Markov in that the transition rates are time-inhomogeneous and stochastic. Numerical solutions to the model provide insight into the effect of aging on component reliability.

  10. 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.

  11. 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.

  12. N reactor level III probabilistic risk assessment using NUREG-1150 methods

    SciTech Connect (OSTI)

    Wang, O.S.; Coles, G.A.; Kelly, J.E.; Powers, T.B.; Rainey, T.E.; Zentner, M.D. (Westinghouse Hanford Co., Richland, WA (US)); Wyss, G.D.; Kunsman, D.M.; Miller, L.A.; Wheeler, T.A.; Sprung, J.L.; Camp, A.L. (Sandia National Lab., Albuquerque, NM (US))

    1991-11-01T23:59:59.000Z

    This paper reports that in the late 1980s, a level III probabilistic risk assessment (PRA) was performed for the N Reactor, a U.S. Department of Energy (DOE) production reactor located on the Hanford site in Washington State. The PRA objectives were to assess the risks to the public and to the Hanford on-site workers posed by the operation of the N Reactor, to compare those risks to proposed DOE nuclear safety guidelines, and to identify risk-reduction changes to the plant. State-of-the-art methodology was employed based largely on the methods developed by Sandia National Laboratories for the U.S. Nuclear Regulatory Commission in support of the NUREG-1150 study of five commercial nuclear power plants. The structure of the probabilistic models allowed complex interactions and dependencies between systems to be explicitly considered. Latin hypercube sampling techniques were used to develop uncertainty distribution for the risks associated with postulated core damage events initiated by fire, seismic, and internal events as well as the overall combined risk. The risk results show that the N Reactor meets the proposed DOE nuclear safety guidelines and compares favorably to the commercial nuclear power plants considered in the NUREG-1150 analysis.

  13. A social impact assessment of the floodwater spreading project on the Gareh-Bygone plain in Iran: A causal comparative approach

    SciTech Connect (OSTI)

    Ahmadvand, Mostafa [Department of Agricultural Extension and Education, College of Agriculture, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: Ahmadvand_2000@yahoo.com; Karami, Ezatollah [Department of Agricultural Extension and Education, College of Agriculture, Shiraz University, Shiraz (Iran, Islamic Republic of)], E-mail: ekarami@shirazu.ac.ir

    2009-02-15T23:59:59.000Z

    The purpose of this study was to explore the social impacts of the floodwater spreading project (FWSP) on the Gareh-Bygone plain, Iran. The study was in the form of a causal comparative design, and a triangulation technique was used to collect data including the use of survey data, archival data, and a participatory rural appraisal (PRA). The causal comparative method requires a comparison of villages with and without the FWSP. Therefore, a survey was conducted using stratified random sampling to select 202 households in villages with and without FWSP in the plain. Significant differences were found between the respondents in villages with and without FWSP with regard to social impact criteria. In spite of the project had negative impact on perceived wellbeing, social capital, social structure development; it had positive impact on quality of life, rural and agricultural economic conditions, and conservation of community resources. However, no significant difference was found between women and men regarding the SIA of FWSP in Gareh-Bygone plain. Analysis of the archival data and PRA techniques supported the survey results and demonstrated that the project improved environmental criteria and deteriorated social dimensions.

  14. 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.

  15. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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

  3. 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.

  4. 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.

  5. 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.

  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. 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.

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

    SciTech Connect (OSTI)

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

    2011-03-13T23:59:59.000Z

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

  9. 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.

  10. 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.

  11. Revised accident progression and risk analyses for NUREG-1150

    SciTech Connect (OSTI)

    Gorham-Bergeron, E.D.; Haskin, F.E.; Hora, S.C.

    1987-01-01T23:59:59.000Z

    Preliminary III PRA analyses that support preparation of the Nuclear Regulatory Commission's Reactor Risk Reference Document (NUREG-1150) have been conducted at Sandia National Laboratories for four plants: Surry, Sequoyah, Peach Bottom and Grand Gulf. Brookhaven National Laboratories conducted the analysis for the Zion plant. Review of the preliminary analyses produced comments and criticisms from two committees (Kouts Committee and Kastenberg Committee), from industry, and from a variety of other sources. As a result, the final analyses currently under way at Sandia and Brookhaven will contain several improvements over the preliminary analyses. Of these the most significant improvement is in the methodology used to elicit expert opinion concerning highly uncertain questions about severe accident phenomena. 17 refs., 1 fig., 1 tab.

  12. The risk management implications of NUREG--1150 methods and results

    SciTech Connect (OSTI)

    Camp, A.L.; Maloney, K.J.; Sype, T.T. (Sandia National Labs., Albuquerque, NM (USA))

    1989-09-01T23:59:59.000Z

    This report describes the potential uses of NUREG-1150 and similar Probabilistic Risk Assessments (PRAs) in NRC and industry risk management programs. NUREG-1150 uses state-of-the-art PRA techniques to estimate the risk from five nuclear power plants. The methods and results produced in NUREG-1150 provide a framework within which current risk management strategies can be evaluated, and future risk management programs can be developed and assessed. While the development of plant-specific risk management strategies is beyond the scope of this document, examples of the use of the NUREG-1150 framework for identifying and evaluating risk management options are presented. All phases of risk management from prevention of initiating events though reduction of offsite consequences are discussed, with particular attention given to the early phase of accidents. 14 refs., 9 figs., 28 tabs.

  13. NUREG-1150 risk assessment methodology

    SciTech Connect (OSTI)

    Benjamin, A.S.; Amos, C.N.; Cunningham, M.A.; Murphy, J.A.

    1987-01-01T23:59:59.000Z

    This paper describes the methodology developed in support of the US Nuclear Regulatory Commission's (NCR's) evaluation of severe accident risks in NUREG-1150. After the accident at Three Mile Island, Unit 2, the NRC initiated a sever accident research program to develop an improved understanding of severe accidents and to provide a second technical basis to support regulatory decisions in this area. A key product of this program is NUREG-1150, which provides estimates of risk for several nuclear reactors of different design. The principal technical analyses for NUREG-1150 were performed at Sandia National Labs. under the Severe Accident Risk Reduction Program and the Accident Sequence Evaluation Program. A major aspect of the work was the development of a methodology that improved upon previous full-scale probabilistic risk assessments (PRA) in several areas which are described.

  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. 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.

  16. 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.

  17. 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.

  18. 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.

  19. 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.

  20. 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.

  1. 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.

  2. 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.

  3. 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.

  4. 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.

  5. 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.

  6. 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.

  7. 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.

  8. 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.

  9. 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.

  10. 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.

  11. 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.)

  12. Large break loss of coolant severe accident sequences at the HFIR (High Flux Isotope Reactor)

    SciTech Connect (OSTI)

    Simpson, D.B.; Greene, S.R.

    1990-01-01T23:59:59.000Z

    An assessment of many potential HFIR severe accident phenomena was conducted during the HFIR design effort, and many severe accident mitigating features were designed into the plant. These evaluation typically incorporated a bounding'' or highly conservative analysis approach and employed tools and techniques representative of the state of knowledge in the mid-1960s. Recently, programs to address severe accident issues were initiated at the Oak Ridge National Laboratory (ORNL) to support the HFIR probabilistic risk assessment (PRA) and equipment qualification and accident management studies. This paper presents the results of environment condition calculations conducted to evaluate a response of HFIR's heat exchanger cell environment to a double-ended rupture of a 0.25 m diameter coolant loop downstream of the circulating pump and check valve. The confinement calculations were performed using an atmospheric fission product source for the heat exchanger cell consistent with, but more conservative than that stipulated in Regulatory Guide 1.89. The results of the calculations indicate that the heat exchanger cell atmospheric temperature peaks at 377 K 225 seconds into the transient and then begins decreasing at approximately 1.7 K per minute. 8 refs., 5 figs.

  13. On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

    SciTech Connect (OSTI)

    Freels, J.D.

    1993-01-01T23:59:59.000Z

    This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as the code''). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed.

  14. On RELAP5-simulated High Flux Isotope Reactor reactivity transients: Code change and application

    SciTech Connect (OSTI)

    Freels, J.D.

    1993-07-01T23:59:59.000Z

    This paper presents a new and innovative application for the RELAP5 code (hereafter referred to as ``the code``). The code has been used to simulate several transients associated with the (presently) draft version of the High-Flux Isotope Reactor (HFIR) updated safety analysis report (SAR). This paper investigates those thermal-hydraulic transients induced by nuclear reactivity changes. A major goal of the work was to use an existing RELAP5 HFIR model for consistency with other thermal-hydraulic transient analyses of the SAR. To achieve this goal, it was necessary to incorporate a new self-contained point kinetics solver into the code because of a deficiency in the point-kinetics reactivity model of the Mod 2.5 version of the code. The model was benchmarked against previously analyzed (known) transients. Given this new code, four event categories defined by the HFIR probabilistic risk assessment (PRA) were analyzed: (in ascending order of severity) a cold-loop pump start; run-away shim-regulating control cylinder and safety plate withdrawal; control cylinder ejection; and generation of an optimum void in the target region. All transients are discussed. Results of the bounding incredible event transient, the target region optimum void, are shown. Future plans for RELAP5 HFIR applications and recommendations for code improvements are also discussed.

  15. 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.

  16. A pilot application of risk-informed methods to establish inservice inspection priorities for nuclear components at Surry Unit 1 Nuclear Power Station. Revision 1

    SciTech Connect (OSTI)

    Vo, T.V.; Phan, H.K.; Gore, B.F.; Simonen, F.A.; Doctor, S.R. [Pacific Northwest National Lab., Richland, WA (United States)

    1997-02-01T23:59:59.000Z

    As part of the Nondestructive Evaluation Reliability Program sponsored by the US Nuclear Regulatory Commission, the Pacific Northwest National Laboratory has developed risk-informed approaches for inservice inspection plans of nuclear power plants. This method uses probabilistic risk assessment (PRA) results to identify and prioritize the most risk-important components for inspection. The Surry Nuclear Power Station Unit 1 was selected for pilot application of this methodology. This report, which incorporates more recent plant-specific information and improved risk-informed methodology and tools, is Revision 1 of the earlier report (NUREG/CR-6181). The methodology discussed in the original report is no longer current and a preferred methodology is presented in this Revision. This report, NUREG/CR-6181, Rev. 1, therefore supersedes the earlier NUREG/CR-6181 published in August 1994. The specific systems addressed in this report are the auxiliary feedwater, the low-pressure injection, and the reactor coolant systems. The results provide a risk-informed ranking of components within these systems.

  17. 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

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

    SciTech Connect (OSTI)

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

  19. Fiber scrambling for high-resolution spectrographs. II. A double fiber scrambler for Keck Observatory

    E-Print Network [OSTI]

    Spronck, Julien F P; Kaplan, Zachary; Jurgenson, Colby; Valenti, Jeff; Moriarty, John; Szymkowiak, Andrew E

    2015-01-01T23:59:59.000Z

    We have designed a fiber scrambler as a prototype for the Keck HIRES spectrograph, using double scrambling to stabilize illumination of the spectrometer and a pupil slicer to increase spectral resolution to R = 70,000 with minimal slit losses. We find that the spectral line spread function (SLSF) for the double scrambler observations is 18 times more stable than the SLSF for comparable slit observations and 9 times more stable than the SLSF for a single fiber scrambler that we tested in 2010. For the double scrambler test data, we further reduced the radial velocity scatter from an average of 2.1 m/s to 1.5 m/s after adopting a median description of the stabilized SLSF in our Doppler model. This demonstrates that inaccuracies in modeling the SLSF contribute to the velocity RMS. Imperfect knowledge of the SLSF, rather than stellar jitter, sets the precision floor for chromospherically quiet stars analyzed with the iodine technique using Keck HIRES and other slit-fed spectrometers. It is increasingly common pra...

  20. 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.

  1. 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.

  2. Probabilistic risk assessment of the N Reactor using NUREG-1150 methods

    SciTech Connect (OSTI)

    Wang, O.S.; Coles, G.A.; Zentner, M.D.; Powers, T.B.; Baxter, J.T.

    1989-01-01T23:59:59.000Z

    A level-III probabilistic risk assessment (PRA) has been performed for N Reactor, a US Department of Energy (DOE) reactor located on the Hanford site in Washington state. The methods developed for the US Nuclear Regulatory Commission (NRC) by Sandia National Laboratories (SNL) in support of the NUREG-1150 effort were adapted for the analysis. The objectives of the study are to assess the risks to the public (off-site) and to workers at colocated DOE facilities (on-site) posed by the operation of N reactor, compare those risks to proposed DOE safety guidelines and NRC safety goals, and identify changes to the plant for safety enhancement. This summary is based on results from internally initiated events only. Off-site risk resulted from externally initiated events and on-site risk will be reported at a later date. The entire study is based on best-estimate inputs except that a number of areas, such as source term inventory and progression of metal/water reaction, apply conservative assumptions.

  3. 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.

  4. 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).

  5. Experimental EPR-Steering of Bell-local States

    E-Print Network [OSTI]

    D. J. Saunders; S. J. Jones; H. M. Wiseman; G. J. Pryde

    2009-09-04T23:59:59.000Z

    Entanglement is the defining feature of quantum mechanics, and understanding the phenomenon is essential at the foundational level and for future progress in quantum technology. The concept of steering was introduced in 1935 by Schr\\"odinger as a generalization of the Einstein-Podolsky-Rosen (EPR) paradox. Surprisingly, it has only recently been formalized as a quantum information task with arbitrary bipartite states and measurements, for which the existence of entanglement is necessary but not sufficient. Previous experiments in this area have been restricted to the approach of Reid [PRA 40, 913], which followed the original EPR argument in considering only two different measurement settings per side. Here we implement more than two settings so as to be able to demonstrate experimentally, for the first time, that EPR-steering occurs for mixed entangled states that are Bell-local (that is, which cannot possibly demonstrate Bell-nonlocality). Unlike the case of Bell inequalities, increasing the number of measurement settings beyond two--we use up to six--dramatically increases the robustness of the EPR-steering phenomenon to noise.

  6. 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.

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

    SciTech Connect (OSTI)

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

    2014-08-01T23:59:59.000Z

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

  8. 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)

  9. Effects of improved modeling on best estimate BWR severe accident analysis

    SciTech Connect (OSTI)

    Hyman, C.R.; Ott, L.J.

    1984-01-01T23:59:59.000Z

    Since 1981, ORNL has completed best estimate studies analyzing several dominant BWR accident scenarios. These scenarios were identified by early Probabilistic Risk Assessment (PRA) studies and detailed ORNL analysis complements such studies. In performing these studies, ORNL has used the MARCH code extensively. ORNL investigators have identified several deficiencies in early versions of MARCH with regard to BWR modeling. Some of these deficiencies appear to have been remedied by the most recent release of the code. It is the purpose of this paper to identify several of these deficiencies. All the information presented concerns the degraded core thermal/hydraulic analysis associated with each of the ORNL studies. This includes calculations of the containment response. The period of interest is from the time of permanent core uncovery to the end of the transient. Specific objectives include the determination of the extent of core damage and timing of major events (i.e., onset of Zr/H/sub 2/O reaction, initial clad/fuel melting, loss of control blade structure, etc.). As mentioned previously the major analysis tool used thus far was derived from an early version of MARCH. BWRs have unique features which must be modeled for best estimate severe accident analysis. ORNL has developed and incorporated into its version of MARCH several improved models. These include (1) channel boxes and control blades, (2) SRV actuations, (3) vessel water level, (4) multi-node analysis of in-vessel water inventory, (5) comprehensive hydrogen and water properties package, (6) first order correction to the ideal gas law, and (7) separation of fuel and cladding. Ongoing and future modeling efforts are required. These include (1) detailed modeling for the pressure suppression pool, (2) incorporation of B/sub 4/C/steam reaction models, (3) phenomenological model of corium mass transport, and (4) advanced corium/concrete interaction modeling. 10 references, 17 figures, 1 table.

  10. 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.

  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. 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.

  13. 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.

  14. 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.

  15. 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.

  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. 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.

  18. 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.

  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. 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.

  1. 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.

  2. 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.

  3. 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.