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1

Di usive Kinetic Explicit Schemes for Nonlinear Degenerate Parabolic Systems  

E-Print Network (OSTI)

Di#11;usive Kinetic Explicit Schemes for Nonlinear Degenerate Parabolic Systems #3; D. Aregba parabolic systems. These schemes are based on discrete BGK models where both characteristic velocities. Evje and K.H. Karlsen [15] and of M. Espedal and K.H. Karlsen [14]. For the theory of general parabolic

2

Fire PRA Methods Enhancements  

Science Conference Proceedings (OSTI)

This report describes research on fire probabilistic risk assessment PRA methods. The fire PRA methods presented in this report provide additions, clarifications, and refinements to the methods proposed in 2005 by the Electric Power Research Institute EPRI and the U.S. Nuclear Regulatory Commission NRC in EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities EPRI 1011989/NUREG/CR-6850. The purpose of the current report is to provide the most current, state-of-the-art information in order to supp...

2008-12-23T23:59:59.000Z

3

Development of PRA Qualification Guidance and Curriculum  

Science Conference Proceedings (OSTI)

This interim report presents the status of a project to develop PRA qualification guidance and a curriculum for the training of utility PRA engineers. At this time, a survey has been prepared and sent to some representative utilities, NSSS vendors, PRA contractors, government agencies, and universities to ascertain existing practices. Their responses will be tabulated and analyzed to characterize typical and best practices, and to provide a basis for recommendations. Subsequent work will inventory existi...

2004-12-22T23:59:59.000Z

4

EPRI Guidelines for PRA Data Analysis  

Science Conference Proceedings (OSTI)

One of the primary tasks to support probabilistic risk assessment (PRA) is a thorough and traceable data analysis to provide as much plant-specific history as possible to the PRA modeling, or well-founded industry information where plant-specific data are insufficient or unavailable, and to characterize the uncertainty surrounding this information to allow sensitivity analyses to be performed.Recent efforts have been made by industry and the Nuclear Regulatory Commission (NRC) to ...

2013-12-13T23:59:59.000Z

5

PRA Fundamentals Computer Based Training Module 2.1  

Science Conference Proceedings (OSTI)

This product is a computer based training (CBT) module on Probabilistic Risk Assessment (Module 2 PRA Fundamentals). This product is a computer based training module on Probabilistic Risk Assessment (PRA CBT Module 2 - PRA Fundamentals). The PRA CBT is the second module and it titled PRA Fundamentals. This module is designed to provide a high level intriduction to the fundamentals of PRAs for nuclear power plants. It is the second module in the series. WindowsXP/Vista/Windows 7.

2010-11-30T23:59:59.000Z

6

Certification plan for safety and PRA codes  

Science Conference Proceedings (OSTI)

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.

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

7

Certification plan for safety and PRA codes  

Science Conference Proceedings (OSTI)

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.

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

8

Demonstrate Ames Laboratory capability in Probabilistic Risk Assessment (PRA)  

Science Conference Proceedings (OSTI)

In response to the damage which occurred during the Three Mile Island nuclear accident, the Nuclear Regulatory Commission (NRC) has implemented a Probabilistic Risk Assessment (PRA) program to evaluate the safety of nuclear power facilities during events with a low probability of occurrence. The PRA can be defined as a mathematical technique to identify and rank the importance of event sequences that can lead to a severe nuclear accident. Another PRA application is the evaluation of nuclear containment buildings due to earthquakes. In order to perform a seismic PRA, the two conditional probabilities of ground motion and of structural failure of the different components given a specific earthquake are first studied. The first of these is termed probability of exceedance and the second as seismic fragility analysis. The seismic fragility analysis is then related to the ground motion measured in terms of ``g`` to obtain a plant level fragility curve.

Bluhm, D.; Greimann, L.; Fanous, F.; Challa, R.; Gupta, S.

1993-07-01T23:59:59.000Z

9

CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection (PRA), &  

NLE Websites -- All DOE Office Websites (Extended Search)

CONTACTS FOR INFORMATION MANAGEMENT: CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection (PRA), & Records CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection (PRA), & Records Name Contact Information Area of Responsibility Troy Manigault Phone: 301-903-9926 Email: doerm@hq.doe.gov Director, Records Management Division Ivan King Phone: 202-586-4060 Email: ivan.king@hq.doe.gov Records Management Program (Lead) Tonya Meadows Phone: 301-903-1146 Email: tonya.meadows@hq.doe.gov Forms Management Program (Lead) Christina "Chris" Rouleau Phone: 301-903-6227 Email: Christina.Rouleau@hq.doe.gov Information Collection Management Program (Lead) Deidra "Dee Dee" Wilkinson Phone: 202-586-2398 Email: deidre.wilkinson@hq.doe.gov Records Management Program

10

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

SciTech Connect

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.

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

1990-01-01T23:59:59.000Z

11

EPRI Probabilistic Risk Assessments (PRA) Computer Based Training (CBT) v1.0  

Science Conference Proceedings (OSTI)

This Computer Based Training (CBT) module provides a high level introduction to the fundamentals of Probabilistic Risk Assessment (PRA) and its use in Risk Informed (RI) Regulation. The EPRI Risk and Safety Management (RSM) Program is developing a series of CBT Modules to assist in the socialization of risk technology, more specifically the understanding of the plant specific Probabilistic Risk Assessments (PRA) and risk informed regulation. The series of PRA CBT modules are developed in a hierarchical o...

2010-01-22T23:59:59.000Z

12

Spatially Informed Plant PRA Models for Security Assessment  

SciTech Connect

Traditional risk models can be adapted to evaluate plant response for situations where plant systems and structures are intentionally damaged, such as from sabotage or terrorism. This paper describes a process by which traditional risk models can be spatially informed to analyze the effects of compound and widespread harsh environments through the use of 'damage footprints'. A 'damage footprint' is a spatial map of regions of the plant (zones) where equipment could be physically destroyed or disabled as a direct consequence of an intentional act. The use of 'damage footprints' requires that the basic events from the traditional probabilistic risk assessment (PRA) be spatially transformed so that the failure of individual components can be linked to the destruction of or damage to specific spatial zones within the plant. Given the nature of intentional acts, extensive modifications must be made to the risk models to account for the special nature of the 'initiating events' associated with deliberate adversary actions. Intentional acts might produce harsh environments that in turn could subject components and structures to one or more insults, such as structural, fire, flood, and/or vibration and shock damage. Furthermore, the potential for widespread damage from some of these insults requires an approach that addresses the impacts of these potentially severe insults even when they occur in locations distant from the actual physical location of a component or structure modeled in the traditional PRA. (authors)

Wheeler, Timothy A. [Sandia National Laboratories, PO Box 5800, Albuquerque, NM 87185 (United States); Thomas, Willard [Omicron Safety and Risk Technologies, Inc., 2500 Louisiana Boulevard, Suite 410. Albuquerque, NM 87110 (United States); Thornsbury, Eric [U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001 (United States)

2006-07-01T23:59:59.000Z

13

Level 1 Tornado PRA for the High Flux Beam Reactor  

Science Conference Proceedings (OSTI)

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

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

1994-05-01T23:59:59.000Z

14

Total Risk Approach in Applying PRA to Criticality Safety  

SciTech Connect

As nuclear industry continues marching from an expert-base support to more procedure-base support, it is important to revisit the total risk concept to criticality safety. A key objective of criticality safety is to minimize total criticality accident risk. The purpose of this paper is to assess key constituents of total risk concept pertaining to criticality safety from an operations support perspective and to suggest a risk-informed means of utilizing criticality safety resources for minimizing total risk. A PRA methodology was used to assist this assessment. The criticality accident history was assessed to provide a framework for our evaluation. In supporting operations, the work of criticality safety engineers ranges from knowing the scope and configurations of a proposed operation, performing criticality hazards assessment to derive effective controls, assisting in training operators, response to floor questions, surveillance to ensure implementation of criticality controls, and response to criticality mishaps. In a compliance environment, the resource of criticality safety engineers is increasingly being directed towards tedious documentation effort to meet some regulatory requirements to the effect of weakening the floor support for criticality safety. By applying a fault tree model to identify the major contributors of criticality accidents, a total risk picture is obtained to address relative merits of various actions. Overall, human failure is the key culprit in causing criticality accidents. Factors such as failure to follow procedures, lacks of training, lack of expert support at the floor level etc. are main contributors. Other causes may include lack of effective criticality controls such as inadequate criticality safety evaluation. Not all of the causes are equally important in contributing to criticality mishaps. Applying the limited resources to strengthen the weak links would reduce risk more than continuing emphasis on the strong links of criticality safety support. For example, some compliance failures such as lack of detailed documentation may not be as relevant as the lack of floor support in answering operator's questions during operations. Misuse of resources in reducing lesser causes rather than on major causes of criticality accidents is not risk free without severe consequences. A regulatory mandate without due consideration of total risk may have its opposite effect of increasing the total risk of an accident. A lesson is to be learned here. For regulatory standard/guide development, use of ANS/ANSI standard process, which provides the pedigree of consensus participation, is recommended.

Huang, S T

2005-03-24T23:59:59.000Z

15

Selecting the seismic HRA approach for Savannah River Plant PRA revision 1  

SciTech Connect

The Westinghouse Savannah River Company (WSRC) has prepared a level I probabilistic risk assessment (PRA), Rev. 0 of reactor operations for externally-initiated events including seismic events. The SRS PRA, Rev. 0 Seismic HRA received a critical review that expressed skepticism with the approach used for human reliability analysis because it had not been previously used and accepted in other published PRAs. This report provides a review of published probabilistic risk assessments (PRAs), the associated methodology guidance documents, and the psychological literature to identify parameters important to seismic human reliability analysis (HRA). It also describes a recommended approach for use in the Savannah River Site (SRS) PRA. The SRS seismic event PRA performs HRA to account for the contribution of human errors in the accident sequences. The HRA of human actions during and after a seismic event is an area subject to many uncertainties and involves significant analyst judgment. The approach recommended by this report is based on seismic HRA methods and associated issues and concerns identified from the review of these referenced documents that represent the current state-of-the- art knowledge and acceptance in the seismic HRA field.

Papouchado, K.; Salaymeh, J. [eds.] [Westinghouse Savannah River Co., Aiken, SC (United States); Wingo, H.E.; Benhardt, H.C.; van Buijtenen, C.M.; Mitts, T.M. [Battelle Pacific Northwest Labs., Richland, WA (United States)

1993-10-01T23:59:59.000Z

16

Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Quantification and Analysis Report  

Science Conference Proceedings (OSTI)

Since the U.S. Nuclear Regulatory Commission (NRC) approved dry storage of spent fuel, dry casks stored at U.S. sites have increased significantly in number since the 1980s. This project, a spent fuel cask probabilistic risk assessment (PRA), was designed to obtain insights related to the risks associated with the dry storage of spent fuel.

2003-12-31T23:59:59.000Z

17

Probabilistic Risk Assessment (PRA) of Bolted Storage Casks: Updated Quantification and Analysis Report  

Science Conference Proceedings (OSTI)

Since the 1980s, when the U.S. Nuclear Regulatory Commission (NRC) approved dry storage of spent fuel, dry casks stored at U.S. sites have increased significantly in number. This report and its predecessor — a spent fuel cask probabilistic risk assessment (PRA) (EPRI report 1002877) — are designed to provide insights related to the risks associated with dry storage of spent fuel.

2004-11-30T23:59:59.000Z

18

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

SciTech Connect

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.

Curtis Smith

2013-09-01T23:59:59.000Z

19

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

SciTech Connect

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

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

1990-01-01T23:59:59.000Z

20

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

Science Conference Proceedings (OSTI)

This report presents the development and the first application of generic probabilistic risk assessment (PRA) information for identifying systems and components important to public risk at nuclear power plants lacking plant-specific PRAs. A methodology is presented for using the results of PRAs for similar (surrogate) plants, along with plant-specific information about the plant of interest and the surrogate plants, to infer important failure modes for systems of the plant of interest. This methodology, and the rationale on which it is based, is presented in the context of its application to the Rancho Seco plant. The Rancho Seco plant has been analyzed using PRA information from two surrogate plants. This analysis has been used to guide development of considerable plant-specific information about Rancho Seco systems and components important to minimizing public risk, which is also presented herein.

Gore, B.F.; Huenefeld, J.C.

1987-07-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


21

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

SciTech Connect

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.

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

2011-03-01T23:59:59.000Z

22

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

Science Conference Proceedings (OSTI)

Over a 12 month period, a fire PRA was developed for a DOE facility using the NUREG/CR-6850 EPRI/NRC fire PRA methodology. The fire PRA modeling included calculation of fire severity factors (SFs) and fire non-suppression probabilities (PNS) for each safe shutdown (SSD) component considered in the fire PRA model. The SFs were developed by performing detailed fire modeling through a combination of CFAST fire zone model calculations and Latin Hypercube Sampling (LHS). Component damage times and automatic fire suppression system actuation times calculated in the CFAST LHS analyses were then input to a time-dependent model of fire non-suppression probability. The fire non-suppression probability model is based on the modeling approach outlined in NUREG/CR-6850 and is supplemented with plant specific data. This paper presents the methodology used in the DOE facility fire PRA for modeling fire-induced SSD component failures and includes discussions of modeling techniques for: • Development of time-dependent fire heat release rate profiles (required as input to CFAST), • Calculation of fire severity factors based on CFAST detailed fire modeling, and • Calculation of fire non-suppression probabilities.

Tom Elicson; Bentley Harwood; Jim Bouchard; Heather Lucek

2011-03-01T23:59:59.000Z

23

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

Science Conference Proceedings (OSTI)

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.

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

1990-06-01T23:59:59.000Z

24

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

SciTech Connect

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

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

2010-04-01T23:59:59.000Z

25

EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities Volume 1: Summary & Overview, Volume 2: Detailed Methodology  

Science Conference Proceedings (OSTI)

The Fire Risk Requantification Study has resulted in state-of-the-art methods, tools, and data for a fire probabilistic risk assessment (PRA) for commercial nuclear power plant application. This study was conducted jointly by EPRI and the U.S. Nuclear Regulatory Commission (NRC) Office of Nuclear Regulatory Research (RES) under the terms of an NRC/EPRI Memorandum of Understanding and an accompanying Fire Research Addendum. Industry participants supported demonstration analyses and provided peer review of...

2005-09-19T23:59:59.000Z

26

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

Science Conference Proceedings (OSTI)

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.

Flanagan, G.F.

1990-01-01T23:59:59.000Z

27

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

E-Print Network (OSTI)

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

Paris-Sud XI, Université de

28

Loss of spent fuel pool cooling PRA: Model and results  

Science Conference Proceedings (OSTI)

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

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

1996-09-01T23:59:59.000Z

29

PRA insights applicable to the design of the Broad Applications Test Reactor  

SciTech Connect

Design insights applicable to the design of a new Broad Applications Test Reactor (BATR), being studied at Idaho National Engineering Laboratory, are summarized. Sources of design insights include past probabilistic risk assessments and related studies for department of Energy-owned Class A reactors and for commercial reactors. The report includes a preliminary risk allocation scheme for the BATR.

Khericha, S.T.; Reilly, H.J.

1993-01-01T23:59:59.000Z

30

PoS(PRA2009)051 Our changing view of the blue compact dwarf NGC  

E-Print Network (OSTI)

faint stellar population. The gas dispersion of 9 km s-1 of the outer disk is therefore difficult dispersion of 9 km s-1 which is close to the fixed gas velocity dispersion value of 11 km s-1 used by Leroy-circular velocity components within the gas at inner radii are revealed. The central gas dynamics are con- sistent

Kraan-Korteweg, Renée C.

31

PR-A (Progesterone Receptor A) Transgenic Mice Allow Study of ...  

Research Tools; Developing World; Energy. Energy Efficiency; ... Study of the endocrine basis of breast cancer and cancer of the female reproductive ...

32

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

SciTech Connect

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.

Donald Helton; Hossein Esmaili; Robert Buell

2011-03-01T23:59:59.000Z

33

CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection...  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection (PRA), & Records CONTACTS FOR INFORMATION MANAGEMENT: Forms, Information Collection (PRA), & Records Name Contact...

34

To appear in the Handbook of Markov Chain Monte Carlo Edited by Steve Brooks, Andrew Gelman, Galin Jones, and XiaoLi Meng  

E-Print Network (OSTI)

, University of Toronto Hamiltonian dynamics can be used to produce distant proposals for the Metropolis algorithm, thereby avoiding the slow exploration of the state space that results from the di#usive behaviour distant regions. Despite the large overlap in their application areas, the MCMC and molecular dynamics

Neal, Radford M.

35

Level 2 Probabilistic Risk Assessment: An Advanced Education of Risk Professionals Module  

Science Conference Proceedings (OSTI)

This report provides documentation for Level 2 Probabilistic Risk Assessment (PRA): An Advanced Education of Risk Professionals Module. This new training, offered by the Electric Power Research Institute (EPRI), is designated as PRA 310, Level 2 PRA. It is the first advanced module in the Education of Risk Professionals program. Level 2 PRA builds upon and complements the PRA fundamentals training in the Education of Risk Professionals 100 series.This Level 2 PRA training course consists ...

2013-12-13T23:59:59.000Z

36

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

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

against using PRA for FCFs include excessive cost, lack of methods for analyzing human error, and lack of existing PRA methods. However, structuring scenarios and...

37

Environmental Degradation Nuclear IX-Tours Form  

Science Conference Proceedings (OSTI)

ENVIRONMENTAL DEGRADATION OF MATERIALS IN NUCLEAR POWER SYSTEMS—WATER REACTORS ... PRA Destination Management. 150 Paularino.

38

Global DC Closed Orbit Correction. Experimen.t on. the NSLS X-ray Rin.g  

E-Print Network (OSTI)

~~'i~·ted manuscript has been authored CQntl;.)CiO ( of the U. S. Government er cont;~,-I No. W-31-109-ENG-38.:LJrdingly, the U. S. Government retains a (onexc\\usive, royalty-free license to publish or reproduce the published form of this 1 contribution, or allow others to do SO, for C. Government purposes.

Y. Chung; G. Decker; K. Evans

1992-01-01T23:59:59.000Z

39

A d VA N c E m E N t o f t H E PraCtICEA d VA N c E m E N t o f t H E PraCtICE D I r E C t F r o M C D C e n v I r o n m e n tA L h e A Lt h S e r v I c e S B r A n c h  

E-Print Network (OSTI)

worker and manager food safety training and experience, restaurant and food worker busyness. She helps the branch's Environmental Health Specialists Network (EHS-Net) with the design and implementation of restaurant food safety studies and analysis of eHS-Net Restaurant Food Safety Studies: What

40

Dry Cask Storage Probabilistic Risk Assessment Scoping Study  

Science Conference Proceedings (OSTI)

This report describes and evaluates the current state of risk assessment methodologies applicable to dry cask storage probabilistic risk assessment (PRA) and suggests appropriate approaches for performing the various aspects of a dry cask storage PRA.

2002-03-20T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


41

The uses and benefits of probabilistic risk assessment in nuclear reactor safety  

SciTech Connect

Probabilistic risk assessment (PRA) has proven to be an important tool in the safety assessment of nuclear reactors throughout the world. Decision making with regard to many safety issues has been facilitated by both general insights from and direct application of this technology. Key uses of PRA are discussed and some examples of successful applications are cited. The benefits and limitations of PRA are also discussed as well as the broader outlook for applications of PRA. 9 refs.

Bari, R.A.; Speis, T.P. (Brookhaven National Lab., Upton, NY (USA); Nuclear Regulatory Commission, Washington, DC (USA). Office of Nuclear Regulatory Research)

1989-01-01T23:59:59.000Z

42

Education of Risk Professionals Module 2  

Science Conference Proceedings (OSTI)

This report provides documentation for Module 2 in the Electric Power Research Institute (EPRI) Education of Risk Professionals Probabilistic Risk Assessment (PRA) training program. Module 2 is comprised of PRA 102, Systems Analysis and PRA 102A, Basic PRA Software. Module 2 is the second of six training modules in the Education of Risk Professionals series. Each module is typically one week in length. The entire training program is typically scheduled over the course of 10 months. The PowerPoint slide ...

2010-12-01T23:59:59.000Z

43

A Preliminary Approach to Human Reliability Analysis for External Events with a Focus on Seismic  

Science Conference Proceedings (OSTI)

Substantial research has been performed to-date to develop and improve methods to perform Human Reliability Analysis (HRA) in support of Probabilistic Risk Assessment (PRA). Existing HRA methods, however, were developed primarily for internal events PRA and PRA applications. These methods often contain underlying assumptions that may or may not be applicable to the challenging and new environment created by an ...

2012-12-12T23:59:59.000Z

44

Slide 1  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

1  PRA studies began in the late 1980s  1989, ATR PRA published as a summary report  1991, ATR PRA full report  1994 and 2004 various model changes  2011, Consolidation, update and improvement of previous PRA work  2012/2013, PRA risk monitor implementation 2  The PRA supports the ATR Updated Final Safety Analysis Report (UFSAR)  The PRA provides sufficient information regarding either core or fuel damage (CDF or FDF) to enable ATR personnel to make risk informed decisions  Improved performance in facility operation, testing, maintenance, training, and emergency procedures  Ensure cost-effective approaches and the setting of priorities for plant upgrades and modifications, especially for risk reduction/system improvements

45

A review of NRC staff uses of probabilistic risk assessment  

SciTech Connect

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.

Not Available

1994-03-01T23:59:59.000Z

46

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

Science Conference Proceedings (OSTI)

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

Sasser, K.

1994-06-01T23:59:59.000Z

47

Education of Risk Professionals, Module 5: Large Early Release Frequency and Internal Flooding  

Science Conference Proceedings (OSTI)

This report provides documentation for Module 5 in the Electric Power Research Institute (EPRI) Education of Risk Professionals Probabilistic Risk Assessment (PRA) training program. Module 5 comprises PRA 107, Large Early Release Frequency (LERF), and PRA 108, Internal Flooding Probabilistic Risk Assessment (IFPRA). Module 5 is the fifth of six training modules in the Education of Risk Professionals series. Each module is typically one week in length. The entire training program is typically ...

2013-11-26T23:59:59.000Z

48

Applications of Probabilistic Risk Assessment  

SciTech Connect

This report provides a summary of potential and actual applications of Probabilistic Risk Assessment (PRA) technology and insights. Individual applications are derived from the experiences of a number of US nuclear utilities. This report identifies numerous applications of PRA techniques beyond those typically associated with PRAs. In addition, believing that the future use of PRA techniques should not be limited to those of the past, areas of plant operations, maintenance, and financial resource allocation are discussed. 9 refs., 3 tabs.

Burns, K.J.; Chapman, J.R.; Follen, S.M.; O'Regan, P.J. (Yankee Atomic Electric Co., Bolton, MA (USA))

1991-05-01T23:59:59.000Z

49

Recovery Act National Institute of Standards and Technology ...  

Science Conference Proceedings (OSTI)

... the Paperwork Reduction Act (PRA), unless that collection of information displays a currently valid Office of Management and Budget (OMB) Control ...

2010-10-05T23:59:59.000Z

50

Information Collection Management | Department of Energy  

NLE Websites -- All DOE Office Websites (Extended Search)

Information Collection Management Information Collection Management The Paperwork Reduction Act (PRA) of 1995 requires each Federal agency to seek and obtain approval from the...

51

NVLAP Application Forms  

Science Conference Proceedings (OSTI)

... Energy Efficient Lighting Products, 2013-03-27. ... Paperwork Reduction Act (PRA) requirements approved by the Office of Management and Budget ...

2013-08-19T23:59:59.000Z

52

Information Management | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

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

53

Petri net modeling of fault analysis for probabilistic risk assessment.  

E-Print Network (OSTI)

??Fault trees and event trees have been widely accepted as the modeling strategy to perform Probabilistic Risk Assessment (PRA). However, there are several limitations associated… (more)

Lee, Andrew

2013-01-01T23:59:59.000Z

54

Comments of the Northwest Balancing Authorities  

U.S. Energy Information Administration (EIA)

The EIA has proposed new Form EIA-930 which is being evaluated, along with revisions to existing forms, under the Paperwork Reduction Act (PRA).

55

Water mass transformation due to mixed layer entrainment and mesoscale stirring: In series or parallel?  

E-Print Network (OSTI)

The convergence of advective and di#usive buoyancy flux must match the air-sea buoyancy flux between two outcropping isopycnals. This leads to a diagnostic framework for water mass transformation in which a myriad of di#erent processes can be incorporated under a unifying balance. We review how the diapycnal advection due to ubiquitous mixed layer entrainment can be included in this framework, and we estimate its contribution to the large scale transformation. We also consider how decomposing the flow and buoyancy field into mean, eddy and turbulent parts leads to clarifying the interaction of mixed layer and mesoscale (or sub-mesoscale) eddies in the overall large scale balance. 1.

A. Tandon

2001-01-01T23:59:59.000Z

56

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment September 19, 2012 Presenter: Bentley Harwood, Advanced Test Reactor Nuclear Safety Engineer Battelle Energy Alliance Idaho National Laboratory Topics covered: PRA studies began in the late 1980s 1989, ATR PRA published as a summary report 1991, ATR PRA full report 1994 and 2004 various model changes 2011, Consolidation, update and improvement of previous PRA work 2012/2013, PRA risk monitor implementation Idaho National Laboratory Advanced Test Reactor Probabilistic Risk Assessment More Documents & Publications DOE's Approach to Nuclear Facility Safety Analysis and Management Nuclear Regulatory Commission Handling of Beyond Design Basis Events for

57

Augmenting Probabilistic Risk Assesment with Malevolent Initiators  

SciTech Connect

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.

Curtis Smith; David Schwieder

2011-11-01T23:59:59.000Z

58

A Power and Resolution Adaptive  

E-Print Network (OSTI)

A new power and resolution adaptive flash ADC, named PRA-ADC, is proposed. The PRA-ADC enables exponential power reduction with linear resolution reduction. Unused parallel voltage comparators are switched to standby mode. The voltage comparators consume only the leakage power during the standby mode. The PRA-ADC, capable of operating at 5-bit, 6-bit, 7-bit, and 8-bit precision, dissipates 69 mW at 5-bit and 435 mW at 8-bit. The PRA-ADC was designed and simulated with 0.18 m CMOS technology. The PRA-ADC design is applicable to RF portable communication devices, allowing tighter management of power and e#ciency.

Flash Analog-To-Digital Converter; Jincheol Yoo; Daegyu Lee; Kyusun Choi; Jongsoo Kim

2002-01-01T23:59:59.000Z

59

(Awarded Best Theory Paper!) A Probabilistic Approach to Estimating Computer System Reliability  

Science Conference Proceedings (OSTI)

Probabilistic Risk Assessment (PRA) is a method of estimating system reliability by combining logic models of the ways systems can fail with numerical failure rates. One postulates a failure state and systematically decomposes this state into a combination ...

Robert Apthorpe

2001-12-01T23:59:59.000Z

60

Operator Reliability Experiments Using Power Plant Simulators, Volumes 1-3: Volumes 1-3  

Science Conference Proceedings (OSTI)

Operator reliability experiment (ORE) data, collected on actual operating crews from three PWR and three BWR simulators, enhances understanding of their performance during emergency events. These data form the basis for quantifying human reliability in probabilistic risk assessment (PRA).

1980-08-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


61

Assessing the performance of human-automation collaborative planning systems  

E-Print Network (OSTI)

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

Ryan, Jason C. (Jason Christopher)

2011-01-01T23:59:59.000Z

62

Utility Application Experiences of Probabilistic Risk Assessment Method  

Science Conference Proceedings (OSTI)

Ensuring the reliable delivery of electricity is the primary challenge facing power system operators and planners. This technical report summarizes recent utility experiences of applying EPRI's Probabilistic Risk Assessment (PRA) methodology.

2007-11-28T23:59:59.000Z

63

The use for frequency-consequence curves in future reactor licensing  

E-Print Network (OSTI)

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 ...

Debesse, Laurène

2007-01-01T23:59:59.000Z

64

Applications of the EBR-II Probabilistic Risk Assessment  

Science Conference Proceedings (OSTI)

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future.

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

1993-12-31T23:59:59.000Z

65

Applications of the EBR-II Probabilistic Risk Assessment  

SciTech Connect

A Probabilistic Risk Assessment (PRA) of the Experimental Breeder Reactor 11 (EBR-11), a Department of Energy (DOE) Category A research reactor, has recently been completed at Argonne National Laboratory (ANL), and has been performed with close collaboration between PRA analysts and engineering and operations staff. A product of this Involvement of plant personnel has been a excellent acceptance of the PRA as a tool, which has already resulted In a variety of applications of the EBR-11 PRA. The EBR-11 has been used in support of plant hardware and procedure modifications and In new system design work. A new application in support of the refueling safety analysis will be completed in the near future.

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

1993-01-01T23:59:59.000Z

66

New Program Review and Analysis Office to Improve NNSA's Budgeting...  

National Nuclear Security Administration (NNSA)

in restructuring several major defense acquisition programs. He most recently led the OSD-CAPE review of the B61 Life Extension Program. "With the creation of PR&A, we...

67

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

E-Print Network (OSTI)

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 ...

Minton, Mark A. (Mark Aaron)

2010-01-01T23:59:59.000Z

68

Criteria for assessing the quality of nuclear probabilistic risk assessments  

E-Print Network (OSTI)

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

Zhu, Yingli, 1976-

2004-01-01T23:59:59.000Z

69

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

E-Print Network (OSTI)

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 ...

Dulac, Nicolas, 1978-

2007-01-01T23:59:59.000Z

70

Probabilistic Risk Assessment Compendium of Candidate Consensus Models  

Science Conference Proceedings (OSTI)

This report provides a compendium of candidate consensus models in use in current probabilistic risk assessments (PRAs). The ASME PRA Standard, as modified and endorsed by Regulatory Guide 1.200, establishes that the identification, sensitivity analysis, and documentation of key sources of uncertainties and key assumptions may be reduced in scope if the PRA makes use of consensus models to implement the supporting requirements. As part of the process of treating the uncertainties associated with a risk-...

2006-08-16T23:59:59.000Z

71

Pipe Rupture Frequencies for Internal Flooding Probabilistic Risk Assessments: Revision 3  

Science Conference Proceedings (OSTI)

This report updates a 2010 EPRI report (1021086) on piping system failure rates for use in probabilistic risk assessments (PRAs) involving internal plant flooding and high-energy line breaks (HELBs) and represents the third revision to this pipe failure rate handbook. These failure rate estimates are intended to satisfy requirements of the American Society of Mechanical Engineers (ASME) and American Nuclear Society (ANS) PRA Standard RA-Sa-2009. The estimates also support an EPRI PRA Scope and ...

2013-04-26T23:59:59.000Z

72

Probabilistic Risk Assessment - Insights for Executives  

Science Conference Proceedings (OSTI)

This report describes the development and use of a new Probabilistic Risk Assessment (PRA) training resource, Probabilistic Risk Assessment – Insights for Executives.  It consists of a slide package with speaker notes, for use by EPRI members to expand understanding of PRA and its applications among executives and nuclear power plant leadership teams.BackgroundWhile many technical disciplines in the nuclear industry have gained acceptance and are ...

2013-07-31T23:59:59.000Z

73

Single-Point Vulnerabilities in a Transmission and Distribution System  

Science Conference Proceedings (OSTI)

Probabilistic risk assessment (PRA) tools and modeling techniques can be used to evaluate a variety of complex systems and facilities. This report presents an application of PRA techniques to an electric transmission and distribution (TD) system. The work focuses on the reliability of a small utility system providing electric power to approximately 100,000 customers. It examines the probability of loss of systemwide service as well as loss of power to critical facilities (for example, hospitals). The eva...

2011-04-21T23:59:59.000Z

74

Support System Initiating Events  

Science Conference Proceedings (OSTI)

This report documents current methods to identify and quantify support system initiating events SSIEs used in probabilistic risk assessment PRA. This report updates the guidance provided in an EPRI report published in 2006, Support System Initiating Events: Identification and Quantification Guideline 1013490, and has been developed with input from a broad spectrum of the PRA community. Cooperation with the U.S. Nuclear Regulatory Commission NRC and Idaho National Laboratory INL, under the provisions of t...

2008-12-19T23:59:59.000Z

75

Practical Guidance on the Use of Probabilistic Risk Assessment in Risk-Informed Applications with a Focus on the Treatment of Uncert ainty  

Science Conference Proceedings (OSTI)

BackgroundNeither the Nuclear Regulatory Commission (NRC) nor the industry intends to use probabilistic risk assessment (PRA) as a total replacement for traditional deterministic approaches. PRA is viewed as a complement to the deterministic method. In fact, probabilistic and deterministic methods are acknowledged as extensions of each other rather than as separate and distinct.Both the industry and the NRC are incorporating risk concepts and techniques into ...

2012-12-04T23:59:59.000Z

76

Treatment of Parameter and Modeling Uncertainty for Probabilistic Risk Assessments  

Science Conference Proceedings (OSTI)

Both the industry and the U.S. Nuclear Regulatory Commission (NRC) incorporate risk concepts and techniques into activities for effective risk management. The NRC is using probabilistic risk assessment (PRA) in its regulatory activities in a manner that promotes consistency, predictability, and efficiency in the performance of the NRCs roles of risk manager and protector of public health and safety. The nuclear industry uses PRA to identify and manage risks, as a tool to promote efficient regulatory inte...

2008-12-19T23:59:59.000Z

77

Guideline for the Treatment of Uncertainty in Risk-Informed Applications  

Science Conference Proceedings (OSTI)

In all scientific and engineering endeavors, there exists some level of uncertainty about the outcome. The development and application of a probabilistic risk assessment (PRA) is no exception. The uncertainty in the PRA model and its results manifests itself in several forms, including parametric, modeling, and completeness uncertainty. Various methods exist to both account for uncertainty in the model and evaluate the impact of uncertainty on the outcome of the analysis. However, these methods have gene...

2006-10-09T23:59:59.000Z

78

Investigation of Episodic Flow from Unsaturated Porous Media into a Macropore  

DOE Green Energy (OSTI)

Th e recent literature contains numerous observations of episodic or intermittent fl ow in unsaturated flow systems under both constant fl ux and ponded boundary conditions. Flow systems composed of a heterogeneous porous media, as well as discrete fracture networks, have been cited as examples of systems that can exhibit episodic fl ow. Episodic outfl ow events are significant because relatively large volumes of water can move rapidly through an unsaturated system, carrying water and contaminants to depth greatly ahead of a wetting front predicted by a one-dimensional, gravity-driven diff usive infiltration model. In this study, we model the behavior of water flow through a sand column underlain by an impermeable-walled macropore. Relative permeability and capillary pressure relationships were developed that capture the complex interrelationships between the macropore and the overlying porous media that control fl ow out of the system. The potential for episodic flow is assessed and compared to results of conventional modeling approaches and experimental data from the literature. Model results using coupled matrix–macropore relative permeability and capillary pressure relationships capture the behavior observed in laboratory experiments remarkably well, while simulations using conventional relative permeability and capillary pressure functions fail to capture some of the observed fl ow dynamics. Capturing the rapid downward movement of water suggests that the matrix-macropore capillary pressure and relative permeability functions developed have the potential to improve descriptions of fl ow and transport processes in heterogeneous, variably saturated media.

R. K. Podgorney; J. P. Fairley

2008-02-01T23:59:59.000Z

79

Validation of seismic probabilistic risk assessments of nuclear power plants  

SciTech Connect

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

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

1994-01-01T23:59:59.000Z

80

COMPONENT DEGRADATION SUSCEPTIBILITIES AS THE BASES FOR MODELING REACTOR AGING RISK  

Science Conference Proceedings (OSTI)

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.

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

2010-07-18T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


81

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

Science Conference Proceedings (OSTI)

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.

Groth, Katrina M.; Swiler, Laura Painton

2013-03-01T23:59:59.000Z

82

Peer Review of NRC Standardized Plant Analysis Risk Models  

SciTech Connect

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

Anthony Koonce; James Knudsen; Robert Buell

2011-03-01T23:59:59.000Z

83

SAPHIRE 8 Volume 3 - Users' Guide  

SciTech Connect

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.

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

2011-03-01T23:59:59.000Z

84

ePSA- Standard Self Assessments (SSA), Version 3.2  

Science Conference Proceedings (OSTI)

ePSA Standards Self Assessment (SSA) is a MicrosoftWindows-based program for recording self-assessment data for industry standards on Probabilistic Risk Assessment. ePSA Standards Self Assessment (SSA) is a Microsoft Windows-based program for recording self-assessment data for industry standards on PRA. ePSA SSA helps you manage the job of maintaining your PRA and documenting its compliance with industry standards. This version of ePSA includes templates for the following standards: ASME RA-S-2002 Standa...

2010-05-14T23:59:59.000Z

85

Low Power and Shutdown Risk Assessment Benchmarking Study  

SciTech Connect

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

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

2002-12-15T23:59:59.000Z

86

The Zion integrated safety analysis for NUREG-1150  

SciTech Connect

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.

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

1988-01-01T23:59:59.000Z

87

EPRI/NRC-RES Fire Human Reliability Analysis Guidelines  

Science Conference Proceedings (OSTI)

In 2001, EPRI and Nuclear Regulatory Research (RES) collaborated to improve the state of the art in fire risk assessment to support the new risk-informed environment in fire protection. This project produced a consensus documentNUREG/CR-6850 (EPRI report 1011989), Fire PRA Methodology for Nuclear Power Facilitieswhich addresses fire risk during operations at power plants. NUREG/CR-6850 developed high-level guidance on identifying and including human failure events (HFEs) into the fire PRA (FPRA) and a me...

2009-11-30T23:59:59.000Z

88

Nondestructive Evaluation: Probabilistic Risk Assessment Technical Adequacy Guidance for Risk-Informed In-Service Inspection Programs  

Science Conference Proceedings (OSTI)

Risk-informed methodologies have been developed in order to establish alternative in-service inspection (ISI) requirements that are defined as risk-informed in-service inspection (RI ISI) programs. Plant-specific probabilistic risk assessments (PRAs) are typically used during the RI ISI development process. The ASME PRA Standard (for example, ASME RA Sb 2005) and the U.S. Nuclear Regulatory Commission (NRC) Regulatory Guide (RG) 1.200 R1 and R2 have been issued and provide guidance in determining PRA tec...

2011-07-29T23:59:59.000Z

89

Failure rate data for fusion safety and risk assessment  

SciTech Connect

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

Cadwallader, L.C.

1993-01-01T23:59:59.000Z

90

Failure rate data for fusion safety and risk assessment  

SciTech Connect

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

Cadwallader, L.C.

1993-04-01T23:59:59.000Z

91

Kheshbn No. 147 - Spring 2006 - Journal  

E-Print Network (OSTI)

DK" - .pNT ^KD pra tsu TO iy Diy^a - ." pnt^yDi IOO^KII ,i^ONT ^yoa'^ayao'ra ]x iy 0"Diy oxra iy DIO VNÖ s D'o .p' ? npun piraya piKn pK .f?" rnKO Diy TO -]KA f? yn TO uoKnya /

2006-01-01T23:59:59.000Z

92

Guidelines for Preparing Risk-Informed Graded Quality Assurance Program Implementation Request Submittals  

Science Conference Proceedings (OSTI)

EPRI has assessed the role of probabilistic risk assessment (PRA) in the regulation of nuclear power plant quality assurance programs. This report presents nuclear utilities with one example of a methodology and formatting guidance for developing submittals to the U.S. Nuclear Regulatory Commission (NRC) requesting implementation of risk-informed, performance-based "graded" quality assurance programs.

1998-10-28T23:59:59.000Z

93

Interdisciplinary Institute for Innovation How did Fukushima-Daiichi core  

E-Print Network (OSTI)

reactors can successfully face extreme conditions and conclude that the probability of a nuclear accident event per 50.000 reactor years. As of today, one accounts 14500 operating years of nuclear reactors. PRA models describe how nuclear reactor systems will respond to different initiating events that can

94

A study of demographic embodiments of product recommendation agents in electronic commerce  

Science Conference Proceedings (OSTI)

Product Recommendation Agents (PRAs) and other web-based decision aids are deployed extensively to provide online shoppers with virtual advising services. While the design of PRA's functional features has received a high degree of attention in academic ... Keywords: Anthropomorphic interface, Avatar, Electronic commerce, Ethnicity, Gender, Perceived enjoyment, Perceived usefulness, Product recommendation agent, Social presence

Lingyun Qiu; Izak Benbasat

2010-10-01T23:59:59.000Z

95

Kheshbn No. 121- Spring 1993 - Journal  

E-Print Network (OSTI)

p a Jimyizwn a - K ,x>) o>anp > pjyr ,nynx inax nyn ira ^n x a D^X toxn nyjy> nyn ,t:anp nyœnxnytrV x 71a ,":nix>xnynr paxopto>nio ira . (! 0> anp on" ? tra pr^a piyr »r rx ,

1993-01-01T23:59:59.000Z

96

Nordisk kernesikkerhedsforskning Norrnar kjarnryggisrannsknir  

E-Print Network (OSTI)

and preventive maintenance of systems, components and structures at nuclear facilities during operation-informed completion times with configuration risk management or maintenance rule backstop #12;15 Rapport/Report · 32 intervals but rather the whole maintenance system. · The simple expressions used in PRA are insufficient

97

Criticality Risks During Transportation of Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies.

2006-12-14T23:59:59.000Z

98

Trial Plant Review of an American Nuclear Society External Event Probabilistic Risk Assessment Standard  

Science Conference Proceedings (OSTI)

This study examined a representative set of Seismic Probabilistic Risk Assessments (SPRAs) and Seismic Margin Assessments (SMAs) performed for U.S. nuclear plants and evaluated them against the American Nuclear Society's draft External-Event PRA Methodology Standard for conducting Probabilistic Risk Assessment of external events.

2003-09-22T23:59:59.000Z

99

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

Science Conference Proceedings (OSTI)

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

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

2008-08-01T23:59:59.000Z

100

SAPHIRE 8 Volume 7 - Data Loading  

Science Conference Proceedings (OSTI)

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.

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

2011-03-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


101

Resolving arthropod relationships: Present and future insights from evo-devo studies  

E-Print Network (OSTI)

AF domain in the N-terminus of PR-B (Sartorius et al., 1994). PR-A and PR-B appear to have distinct160s are expressed in primary cultures of rat astrocytes (Grenier et al., 2006). Interestingly of the Golgi apparatus (Grenier et al., 2006). Over-expression and siRNA knockdown experiments using

Popadic', Aleksandar

102

Kheshbn No. 129- Spring 1997 - Journal  

E-Print Network (OSTI)

oaR-io pa ,onnR TT pò oaVyn poRns .p>VR R -un pVRnyaonR pçnaçn im .pRia H pa a^n R pRa poRns poipyaonR nam rm ayn a^is

1997-01-01T23:59:59.000Z

103

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

SciTech Connect

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

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

2006-07-01T23:59:59.000Z

104

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

SciTech Connect

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for ansforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with general information such as installation instructions, starting and stopping the program, and some pointers on how to get around inside the program. Next, database concepts and structure are discussed. Following that discussion are nine sections, one for each of the menu options on the SAPHIRE main menu, wherein the purpose and general capabilities for each option are furnished. Next, the capabilities and limitations of the software are provided.

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

2006-07-01T23:59:59.000Z

105

SAPHIRE 8 Volume 1 - Overview and Summary  

Science Conference Proceedings (OSTI)

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.

C. L. Smith; S. T. Wood

2011-03-01T23:59:59.000Z

106

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

Science Conference Proceedings (OSTI)

The Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) is a software application developed for performing a complete probabilistic risk assessment (PRA) using a personal computer. SAPHIRE is funded by the U.S. Nuclear Regulatory Commission (NRC) and developed by the Idaho National Laboratory (INL). The INL's primary role in this project is that of software developer. However, the INL also plays an important role in technology transfer by interfacing and supporting SAPHIRE users comprised of a wide range of PRA practitioners from the NRC, national laboratories, the private sector, and foreign countries. SAPHIRE can be used to model a complex system’s response to initiating events, quantify associated damage outcome frequencies, and identify important contributors to this damage (Level 1 PRA) and to analyze containment performance during a severe accident and quantify radioactive releases (Level 2 PRA). It can be used for a PRA evaluating a variety of operating conditions, for example, for a nuclear reactor at full power, low power, or at shutdown conditions. Furthermore, SAPHIRE can be used to analyze both internal and external initiating events and has special features for transforming models built for internal event analysis to models for external event analysis. It can also be used in a limited manner to quantify risk in terms of release consequences to both the public and the environment (Level 3 PRA). SAPHIRE includes a separate module called the Graphical Evaluation Module (GEM). GEM provides a highly specialized user interface with SAPHIRE that automates SAPHIRE process steps for evaluating operational events at commercial nuclear power plants. Using GEM, an analyst can estimate the risk associated with operational events in a very efficient and expeditious manner. This reference guide will introduce the SAPHIRE Version 7.0 software. A brief discussion of the purpose and history of the software is included along with general information such as installation instructions, starting and stopping the program, and some pointers on how to get around inside the program. Next, database concepts and structure are discussed. Following that discussion are nine sections, one for each of the menu options on the SAPHIRE main menu, wherein the purpose and general capabilities for each option are furnished. Next, the capabilities and limitations of the software are provided.

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

2008-08-01T23:59:59.000Z

107

Slide 1  

NLE Websites -- All DOE Office Websites (Extended Search)

and Qualification of the Methane and Qualification of the Methane Hydrate Resource Potential Associated with the Barrow Gas Fields Kick-off Meeting Morgantown, WV January 9, 2007 Tom Walsh Presentation Outline Overview Team Project Objectives Scope of Work Schedule,Milestones, Deliverables Overview * Last Comprehensive Reservoir Study in 1991-Glenn and Allen * Postulated Presence of Methane Hydrate * Material Balance Models for East Barrow and Walakpa Fields Lend Support to Possible MH Recharge * Potential Significant Impact on Local Resource * Excellent Laboratory for MH Research Location of Study Barrow Gas Fields Participants DOE-NETL NSB PRA UAF Advisory Committee Tim Collett, Chet Paris, Bob Hunter, Bob Swenson, Shirish Patil, Richard Glenn DOE/NETL COR Robert Vagnetti NSB Project Manager Kent Grinage PRA Principal Investigator

108

Risk contribution from low power and shutdown of a pressurized water reactor  

Science Conference Proceedings (OSTI)

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 operation 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 PRA for traditional internal events and a Level 1 PRA on fire, flooding, and seismically induced core damage sequences. 12 refs., 7 tabs.

Chu, T.L.; Pratt, W.T.

1997-04-01T23:59:59.000Z

109

Structural reliability analysis and seismic risk assessment  

SciTech Connect

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

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

1984-01-01T23:59:59.000Z

110

Information Management and Supporting Documentation | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Information Management and Supporting Documentation Information Management and Supporting Documentation Information Management and Supporting Documentation 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 directed to 10 or more persons (including operations of Government-owned, contractor-operated facilities). Under the PRA, OMB approval for each information collection instrument can last a maximum of 3 years. This site provides information about the Paperwork Reduction Act's requirements and guidance in fulfilling those requirements. DOE's Chief Information Officer (CIO) is the Senior Official responsible for DOE's compliance with the Paperwork Reduction Act. Office of Management and

111

Step by Step Instructions  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Step by Step Instructions Step by Step Instructions For Completing An Information Collection Request 1. A determination must be made if a Federal entity has an Information Collection Request (ICR). To assist in making that determination, the Paperwork Reduction Act (PRA)states the following: The PRA requires each Federal agency to seek and obtain Office of Management and Budget (OMB) approval before undertaking a collection of information directed to ten or more people of the general public, including federal contractors, or continuing a collection for which the OMB approval and validity of the OMB control number are about to expire. 2. Once it's been determined that a program has an ICR, the program works with their Headquarters Point of Contact (POC) and prepare a 60-day Federal

112

Site Screening, Site Selection,  

NLE Websites -- All DOE Office Websites (Extended Search)

BeSt PraCtICeS for: BeSt PraCtICeS for: DRAFT Edition Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference therein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The

113

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11 Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11 During the 580th meeting of the Advisory Committee on Reactor Safeguards (ACRS), February 10-12, 2011, we reviewed the staff's white paper, "A Comparison of Integrated Safety Analysis and Probabilistic Risk Assessment." Our Radiation Protection and Nuclear Materials Subcommittee also reviewed this matter during a meeting on January 11, 2011. During these meetings we met with representatives of the NRC staff and the Nuclear Energy Institute. We also had the benefit of the documents referenced. Comparison of Intergrated Safety Analysis (ISA) and Probabilistic Risk

114

Information Collection Management | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Guidance » Information Collection Management Guidance » Information Collection Management Information Collection Management 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 directed to 10 or more persons (including operations of Government-owned, contractor-operated facilities). Under the PRA, OMB approval for each information collection instrument can last a maximum of 3 years. This site provides information about the Paperwork Reduction Act's requirements and guidance in fulfilling those requirements. DOE's Chief Information Officer (CIO) is the Senior Official responsible for DOE's compliance with the Paperwork Reduction Act. Office of Management and

115

PAPERWORK REDUCTION ACT OF 1995  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

PAPERWORK REDUCTION ACT PAPERWORK REDUCTION ACT OF 1995 U. S. DEPARTMENT OF ENERGY INFORMATION COLLECTION MANAGEMENT PROGRAM Chris Rouleau, PRA Officer Records Management Division Office of the Associate Chief Information Officer for IT Planning, Architecture and E-Government Office of the Chief Information Officer Office of the Chief Information Officer 2/16/2010 2 TOPICS  Paperwork Reduction Act (PRA) of 1995 - Law  Paperwork Reduction Act - Overview  Information Collection Requests (ICRs)  Information Collection Request Associated with A Notice of Proposed Rule Making  Program Points of Contacts  Information Collection Clearance Managers  Information Collection Requests Checklist  Drivers  Annual Information Collection Budget  Summary of What To Do  Summary of What NOT

116

Public Outreach  

NLE Websites -- All DOE Office Websites (Extended Search)

BESt PraCtiCES for: BESt PraCtiCES for: First Edition Disclaimer This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference therein to any specific commercial product, process, or service by trade name, trademark, manufacturer, or otherwise does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views

117

The Accident Sequence Precursor program: Methods improvements and current results  

Science Conference Proceedings (OSTI)

Changes in the US NRC Accident Sequence Precursor program methods since the initial program evaluations of 1969-81 operational events are described, along with insights from the review of 1984-85 events. For 1984-85, the number of significant precursors was consistent with the number observed in 1980-81, dominant sequences associated with significant events were reasonably consistent with PRA estimates for BWRs, but lacked the contribution due to small-break LOCAs previously observed and predicted in PWRs, and the frequency of initiating events and non-recoverable system failures exhibited some reduction compared to 1980-81. Operational events which provide information concerning additional PRA modeling needs are also described.

Minarick, J.W.; Manning, F.M.; Harris, J.D.

1987-01-01T23:59:59.000Z

118

Modelling steam explosions for the Savannah River Site reactor probabilistic risk assessment. Revision 1  

Science Conference Proceedings (OSTI)

As part of the probabilistic risk assessment (PRA) for the Savannah River Site (SRS) reactors, a theoretical model has been used to evaluate the magnitude of steam explosions that could occur during postulated severe accidents at the plant. The model predicts pressure, steam generation and mechanical work during the explosion phase of the interaction. The model is applied to two hypothetical events illustrating typical small-scale and large-scale explosions found in the PRA. These two examples show that yield predictions may range from megajoules to gigajoules for kinetic energy, 10 to 1000 kg for steam generated, and 10 to 1000 atm for peak explosion-zone pressures. A brief study is made to characterize the sensitivity of kinetic energy yield to initial fuel mass and fuel temperature. Explosion kinetic energy increases linearly in proportion to fuel mass, but displays a non-linear dependence on fuel temperature over parameter values of interest.

Vonderfecht, B.E. [Westinghouse Savannah River Co., Aiken, SC (United States); Smith, D.C. [Science Applications International Corp., Albuquerque, NM (United States)

1992-12-31T23:59:59.000Z

119

Modelling steam explosions for the Savannah River Site reactor probabilistic risk assessment  

Science Conference Proceedings (OSTI)

As part of the probabilistic risk assessment (PRA) for the Savannah River Site (SRS) reactors, a theoretical model has been used to evaluate the magnitude of steam explosions that could occur during postulated severe accidents at the plant. The model predicts pressure, steam generation and mechanical work during the explosion phase of the interaction. The model is applied to two hypothetical events illustrating typical small-scale and large-scale explosions found in the PRA. These two examples show that yield predictions may range from megajoules to gigajoules for kinetic energy, 10 to 1000 kg for steam generated, and 10 to 1000 atm for peak explosion-zone pressures. A brief study is made to characterize the sensitivity of kinetic energy yield to initial fuel mass and fuel temperature. Explosion kinetic energy increases linearly in proportion to fuel mass, but displays a non-linear dependence on fuel temperature over parameter values of interest.

Vonderfecht, B.E. (Westinghouse Savannah River Co., Aiken, SC (United States)); Smith, D.C. (Science Applications International Corp., Albuquerque, NM (United States))

1992-01-01T23:59:59.000Z

120

SAPHIRE 8 Software Quality Assurance Plan  

Science Conference Proceedings (OSTI)

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.

Curtis Smith

2010-02-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


121

High Pressure Coolant Injection (HPCI) System Risk-Based Inspection Guide for Browns Ferry Nuclear Power Station  

Science Conference Proceedings (OSTI)

The High Pressure Coolant Injection (HPCI) system has been examined from a risk perspective. A System Risk-Based Inspection Guide (S-RIG) has been developed as an aid to HPCI system inspections at the Browns Ferry Nuclear Power Plant, Units 1, 2 and 3. The role of. the HPCI system in mitigating accidents is discussed in this S-RIG, along with insights on identified risk-based failure modes which could prevent proper operation of the system. The S-RIG provides a review of industry-wide operating experience, including plant-specific illustrative examples to augment the PRA and operational considerations in identifying a catalogue of basic PRA failure modes for the HPCI system. It is designed to be used as a reference for routine inspections, self-initiated safety system functional inspections (SSFIs), and the evaluation of risk significance of component failures at the nuclear power plant.

Wong, S.; DiBiasio, A.; Gunther, W. [Brookhaven National Lab., Upton, NY (United States)

1993-09-01T23:59:59.000Z

122

The Updated Fire Events Database: Description of Content and Fire Event Classification Guidance  

Science Conference Proceedings (OSTI)

This report provides a description of the updated and enhanced Fire Events Database (FEDB) developed by the Electric Power Research Institute (EPRI) in cooperation with the U.S. Nuclear Regulatory Commission (NRC). The FEDB is the principal source of fire incident data for use in fire probabilistic risk assessments (FPRAs) as described in EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities (EPRI report 1011989 and NUREG/CR-6850). It provides a comprehensive and consolidated ...

2013-07-22T23:59:59.000Z

123

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

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.

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

2012-09-20T23:59:59.000Z

124

HRA Calculator Version 4.2  

Science Conference Proceedings (OSTI)

HRA Calculator analyzes and calculates human error probabilities in support of probabilistic risk assessments. HRA Calculator takes a 8220toolboxapproach that uses a variety of HRA methods. The PRA Tools / HRA Calculator User Group was formed in 2000 to address the industryneed for HRA tools and to encourage consistency in HRA results. Version 4.2 adds value by expanding the HRA Calculator methods applied, overcoming past limitations on particular parameters, improving the dependency analysis features, ...

2010-11-19T23:59:59.000Z

125

HRA Calculator 4.1.1, Human Reliability Analysis  

Science Conference Proceedings (OSTI)

HRA Calculator analyzes and calculates human error probabilities in support of probabilistic risk assessments. HRA Calculator takes a 8220toolboxapproach that uses a variety of HRA methods. The PRA Tools / HRA Calculator User Group was formed in 2000 to address the industryneed for HRA tools and to encourage consistency in HRA results. Version 4.1.1 adds value by expanding the HRA Calculator methods applied, overcoming past limitations on particular parameters, improving the dependency analysis features...

2009-10-29T23:59:59.000Z

126

Field Evaluation of the Comanagement of Utility Low-Volume Wastes With High-Volume Coal Combustion By-Products: CL Site  

Science Conference Proceedings (OSTI)

This report presents the results of a field study of comanagement of coal combustion by-products at a utility disposal impoundment in the southeastern United States. The study was part of a multiyear effort by the Electric Power Research Institute (EPRI), in cooperation with the Utility Solid Waste Activities Group (USWAG) and individual utility companies, to characterize utility comanagement practices and collect and analyze a comprehensive set of data pertinent to the environmental effects of those pra...

1997-12-09T23:59:59.000Z

127

Human Reliability Analysis (HRA) Calculator Version 4.21  

Science Conference Proceedings (OSTI)

HRA Calculator analyzes and calculates human error probabilities in support of probabilistic risk assessments. HRA Calculator takes a 8220toolboxapproach that uses a variety of HRA methods. The PRA Tools / HRA Calculator User Group was formed in 2000 to address the industryneed for HRA tools and to encourage consistency in HRA results. Version 4.21 adds value by expanding the HRA Calculator methods applied, overcoming past limitations on particular parameters, improving the dependency analysis features,...

2011-06-07T23:59:59.000Z

128

HRA Calculator, Version 4.21 DEMO  

Science Conference Proceedings (OSTI)

HRA Calculator analyzes and calculates human error probabilities in support of probabilistic risk assessments. HRA Calculator takes a “toolbox” approach that uses a variety of HRA methods. The PRA Tools / HRA Calculator User Group was formed in 2000 to address the industry’s need for HRA tools and to encourage consistency in HRA results. Version 4.21 adds value by expanding the HRA Calculator methods applied, overcoming past limitations on particular parameters, improving the ...

2013-03-07T23:59:59.000Z

129

Identification of External Hazards for Analysis in Probabilistic Risk Assessment  

Science Conference Proceedings (OSTI)

This document reports on the assessment of current practices related to the identification of external events (hazards) that can potentially affect the safety of nuclear power plants and provides recommendations on the screening criteria used to perform this identification process. The identification process is intended for use by individual plants, and the identified external events are appropriate candidates for evaluation using probabilistic risk assessment (PRA). One of the outcomes of an external ev...

2011-12-07T23:59:59.000Z

130

Education of Risk Professionals Module 1  

Science Conference Proceedings (OSTI)

This report provides documentation for Module 1 in the Electric Power Research Institute EPRI) Education of Risk Professionals Probabilistic Risk Assessment PRA) training program. Module 1 is the first of six training modules in the Education of Risk Professionals series. Each module is one week in length, and the entire training program is typically scheduled over 10 months. Accompanying this report are the Microsoft PowerPoint slide presentations for Module 1, which contain speaker notes that offer de...

2009-12-23T23:59:59.000Z

131

A REVIEW OF SOFTWARE-INDUCED FAILURE EXPERIENCE.  

SciTech Connect

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

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

2006-09-01T23:59:59.000Z

132

EPRI/NRC-RES Fire Human Reliability Analysis Guidelines  

Science Conference Proceedings (OSTI)

In 2001, the Electric Power Research Institute (EPRI) and the U.S. Nuclear Regulatory Commission's Office of Nuclear Regulatory Research (NRC-RES), operating under a Memorandum of Understanding (MOU), collaborated to improve the state-of-the-art in fire risk assessment to support the new risk-informed environment in fire protection. This project produced a consensus documentNUREG/CR-6850 (EPRI report 1011989), Fire PRA Methodology for Nuclear Power Facilitieswhich addresses fire risk during operations at...

2012-07-24T23:59:59.000Z

133

Risk-Managed Technical Specifications (RMTS) Guidelines  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) has assessed the role of probabilistic risk assessment (PRA) in the regulation of nuclear power station Technical Specifications. This report presents nuclear utilities with a framework and associated general guidance for implementing Risk-Managed Technical Specifications (RMTS) as a partial replacement for existing Technical Specifications. This report was prepared for EPRI with extensive technical input and review by the Nuclear Energy Institute (NEI) Risk-I...

2006-12-01T23:59:59.000Z

134

Risk-Managed Technical Specifications (RMTS) Guidelines  

Science Conference Proceedings (OSTI)

EPRI has assessed the role of probabilistic risk assessment (PRA) in the regulation of nuclear power plant technical specifications. This report presents nuclear utilities with a framework and associated general guidance for implementing risk managed technical specifications (RMTS) as a partial replacement of existing technical specifications. This report was prepared for EPRI with extensive technical input and review by the Nuclear Energy Institute (NEI) Risk-Informed Technical Specifications Task Force...

2005-12-01T23:59:59.000Z

135

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

SciTech Connect

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)

Li, Yunlong; Yilmaz, Fatma; Bedell, Loys [Entergy Nuclear South (United States)

2006-07-01T23:59:59.000Z

136

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

Science Conference Proceedings (OSTI)

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

Brumburgh, G.

1994-08-31T23:59:59.000Z

137

Risk-Managed Technical Specifications (RMTS) Guidelines: Technical Update to EPRI Interim Development Report 1002965  

Science Conference Proceedings (OSTI)

The Electric Power Research Institute (EPRI) has assessed the role of probabilistic risk assessment (PRA) in the regulation of nuclear power plant technical specifications. This report presents nuclear utilities with one example of a technical framework and associated general guidance for implementation of risk-managed technical specifications (RMTS) as a partial replacement of existing conventional plant technical specifications. This report was prepared by EPRI and the Westinghouse Owners Group (WOG) f...

2004-12-11T23:59:59.000Z

138

Program on Technology Innovation: Education of Risk Professionals  

Science Conference Proceedings (OSTI)

EPRI is sponsoring a series of training courses designed to develop the next generation of risk professionals for the nuclear industry. The Education of Risk Professionals project provides the necessary formal training to utility personnel familiar with the operation of their respective nuclear power plants. The formal training will be followed by utility mentoring of students and final signoff of the various elements of the plant-specific probabilistic risk assessment (PRA). Graduates of the training ca...

2007-12-12T23:59:59.000Z

139

Joint Assessment of Cable Damage and Quantification of Effects from Fire (JACQUE-FIRE)  

Science Conference Proceedings (OSTI)

Volume 1 of this report documents the results of a Phenomena Identification and Ranking Table (PIRT) exercise that was undertaken on fire-induced electrical circuit failures that may occur in nuclear power plants when cables are damaged by fires.  Volume 2 documents the PRA expert elicitation results and will include the best estimate conditional probabilities of hot short-induced spurious operations of control circuits, given fire damage to associated cables.  This program was sponsored ...

2012-10-31T23:59:59.000Z

140

MAAP5 Simulation of Accidents at Fukushima Dai-ichi Units 1, 2, and 3  

Science Conference Proceedings (OSTI)

The original MAAP4 code functional design specification (circa 1989) was defined to address the full extent of degraded core accidents with the potential for reflooding of a badly damaged core. It was intended to support probabilistic risk assessment (PRA) and severe accident management guideline (SAMG) applications that previously were limited by the relatively rudimentary design for MAAP3.0B, the predecessor code.The accidents at Fukushima Dai-ichi Units 1, 2, and 3 prompted a ...

2013-02-23T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


141

Emerging Technology and Architecture Approaches for Plug-in Electric Vehicles to Smart Grid Connectivity  

Science Conference Proceedings (OSTI)

This report provides an overview of the latest advances in technologies evolving to facilitate plug-in electric vehicles (PEVs) to Smart Grid integration. It reiterates applicable requirements based on fundamental principles as well as provides a status on the evolving relevant standards space. Multiple technological approaches are presented, compared, and contrasted; and an update on the status of each is provided. The document concludes with early recommendations for utility and automotive industry pra...

2011-12-21T23:59:59.000Z

142

Criticality Risks During Transportation of Spent Nuclear Fuel  

Science Conference Proceedings (OSTI)

This report presents a best-estimate probabilistic risk assessment (PRA) to quantify the frequency of criticality accidents during railroad transportation of spent nuclear fuel casks. The assessment is of sufficient detail to enable full scrutiny of the model logic and the basis for each quantitative parameter contributing to criticality accident scenario frequencies. The report takes into account the results of a 2007 peer review of the initial version of this probabilistic risk assessment, which was pu...

2008-12-10T23:59:59.000Z

143

Risk management activities at the DOE Class A reactor facilities  

Science Conference Proceedings (OSTI)

The probabilistic risk assessment (PRA) and risk management group of the Association for Excellence in Reactor Operation (AERO) develops risk management initiatives and standards to improve operation and increase safety of the DOE Class A reactor facilities. Principal risk management applications that have been implemented at each facility are reviewed. The status of a program to develop guidelines for risk management programs at reactor facilities is presented.

Sharp, D.A. [Westinghouse Savannah River Co., Aiken, SC (United States); Hill, D.J. [Argonne National Lab., IL (United States); Linn, M.A. [Oak Ridge National Lab., TN (United States); Atkinson, S.A. [EG and G Idaho, Inc., Idaho Falls, ID (United States); Hu, J.P. [Brookhaven National Lab., Upton, NY (United States)

1993-12-31T23:59:59.000Z

144

Modeling of Digital Instrumentation and Control in Nuclear Power Plant Probabilistic Risk Assessments  

Science Conference Proceedings (OSTI)

Probabilistic risk assessment (PRA) typically models hardware components in terms of their failure probability and the effects that any given component failure has on the system it resides in. Digital systems, which include both hardware and software, bring new modeling challenges: determination of the appropriate level of detail to use in the logic models, and estimation of failure rates for software components. Failure probabilities for hardware are typically based on operating experience with componen...

2012-07-30T23:59:59.000Z

145

Nuclear Asset Management Database: Phase 2: Prototype Long-term Asset Management Database (LAMDA)  

Science Conference Proceedings (OSTI)

EPRI members engaged in nuclear asset management (NAM) and life cycle management (LCM) view quality equipment reliability and cost data as one of the highest priority needs in a market-driven industry, but less data are available for equipment important to generation than for safety equipment addressed in probabilistic risk assessment (PRA). This interim report describes the second phase of development of the Long-term Asset Management Database (LAMDA).

2004-12-21T23:59:59.000Z

146

Spent Fuel Pool Risk Assessment Integration Framework (Mark I and II BWRs) and Pilot Plant Application  

Science Conference Proceedings (OSTI)

This report documents the development and pilot application of a generic framework and methodology for conducting a probabilistic risk assessment (PRA) for spent fuel pools at BWR plants with Mark I or II containment designs. A key aspect of the study is the consideration of potential synergistic relationships between adverse conditions in the reactor and the spent fuel pool.BackgroundUsed nuclear fuel from the operation of nuclear power plants is typically ...

2013-05-01T23:59:59.000Z

147

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect

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.

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

1994-12-31T23:59:59.000Z

148

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

SciTech Connect

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. The 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. Maine Yankee was selected as one of a series of plants for study. ne 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 Maine Yankee plant.

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

1992-10-01T23:59:59.000Z

149

Cardiorenal-endocrine dynamics during and following volume expansion  

SciTech Connect

The relationship between atrial pressure, atrial natriuretic peptide (ANP), the renin-angiotensin-aldosterone system, and renal hemodynamic and excretory function was examined during and following acute 10% body weight saline volume expansion and measurements were made at 3.3, 6.6, and 10% body weight volume expansion in pentobarbital anesthetized dogs. Right atrial pressure (RAP), pulmonary capillary wedge pressure (PCWP), fractional excretion of Na (FE/sub Na/), and ANP all increased in parallel during volume expansion. Plasma renin activity (PRA) and aldosterone decreased in parallel during 10% volume expansion. ANP, PRA and aldosterone were measured by radioimmunoassay. Following 10% volume expansion, saline was infused at the peak urine flow rate to maintain peak volume expansion. Despite continued saline infusion, RAP, PCWP, and ANP decreased in parallel. In contrast, FE/sub Na/ remained increased, and aldosterone and PRA remained depressed. These studies demonstrate that atrial pressures, ANP, and FE/sub Na/ increase in parallel during volume expansion; this suggests a role for ANP in modulating acute atrial volume overload. During stable volume expansion periods, however, despite a decrease in ANP levels, Na excretion remains elevated, suggesting that non-ANP mechanisms may be important in maintaining natriuresis during stable volume expansion.

Zimmerman, R.S.; Edwards, B.S.; Schwab, T.R.; Heublein, D.M.; Burnett, J.C. Jr.

1987-02-01T23:59:59.000Z

150

ONCORHYNCHUS MYKISS (WALBAUM) IN FRESH WATER AND AFTER SHORT-TERM EXPOSURE TO SEA WATER  

E-Print Network (OSTI)

Freshwater Atlantic salmon (Salmo salar L.) smolts were abruptly transferred to sea water in May and over 3 days blood plasma ion concentrations were determined together with atrial natriuretic peptide (ANP) and plasma renin activity (PRA) using antibodies raised against human ANP and angiotensin I. Blood plasma Na + and Cl ~ levels in smolts increased and, between 24 and 72 h, PRA increased significantly to O.Qngml"^ " 1, while there was a gradual rise in ANP levels to lOpmoll " 1 at 72 h. Similar measurements were made on parr transferred to sea water in September; in these fish Na + and Cl ~ levels increased as in smolts, PRA remained unchanged at about 0.6ngml ~ 1 h ~ 1 and ANP levels increased significantly to about 20pmoir ' at 24 and 72 h. After 2h in sea water parr showed wide variability in ANP levels, in keeping with circulatory stress, hypoxia and increased atrial stretching. Parr transferred to sea water in December showed low drinking rates of 1.95 ml kg " 1 h " 1, even after 20 days, compared to a

Salmon Salmo; Salar L.; Rainbow Trout; N. F. Smith; F. B. Eddyt; A. D. Struthers; C. Talbot

1991-01-01T23:59:59.000Z

151

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

Science Conference Proceedings (OSTI)

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.

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

152

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

SciTech Connect

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.

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

2013-05-01T23:59:59.000Z

153

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

SciTech Connect

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.

Zhegang Ma

2013-09-01T23:59:59.000Z

154

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

SciTech Connect

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.

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

155

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

Science Conference Proceedings (OSTI)

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.

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

156

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

SciTech Connect

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

Golding, D.; White, A. [Clark Univ., Worcester, MA (United States). Center for Technology, Environment, and Development

1990-12-01T23:59:59.000Z

157

Experimental Study of Parametric Dependence of Electron-scale Turbulence in a Spherical Tokamak  

Science Conference Proceedings (OSTI)

Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak eXperiment (NSTX), electron-scale density ?uctuations are studied with a novel tangen- tial microwave scattering system with high radial resolution of ±2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an ELM event, where we found the ?rst clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, cou- pled with linear gyro-kinetic calculations, leads to the identi?cation of the observed instability as toroidal Electron Temperature Gradient (ETG) modes. It is observed that longer wavelength ETG modes, k??s electron temperature and k? is the wavenumber perpendicular to local equilibrium magnetic ?eld), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal di?usivity. Comparisons with nonlinear ETG gyrokinetic simulations shows ETG turbulence may be able to explain the experimental electron heat ?ux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal ?eld, so that electron gyroradius (?e ), electron beta (?e ) and safety factor (q95 ) are kept approximately constant. More than a factor of two change in electron collisionality, ? ?e, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as ? ?e is decreased in this collisonality scan. However, both linear and nonlinear simulations show no or weak dependence with the electron-ion collision frequency, ? e/i . Instead, other equilibrium parameters (safety factor, electron density gradient, for example) a?ect ETG linear growth rate and electron thermal transport more than ? e/i does. Furthermore, electron heat ?ux predicted by the simulations is found to have an order-of-magnitude spatial variation in the experimental mea- surement region and is also found to be much smaller than experimental levels except at one radial location we evaluated. The predicted electron heat ?ux is shown to be strongly anti-correlated with density gradient which varies for a factor of three in the measurement region, which is in agreement with the density gradient dependence study reported in this paper.

Ren, Y; Kaye, S M; Mazzucato, E; Bell, R E; Diallo, A; Domier, C W; LeBlanc, B P; Lee, K C; Smith, D R

2012-05-23T23:59:59.000Z

158

Experimental Study of Parametric Dependence of Electron-scale Turbulence in a Spherical Tokamak  

SciTech Connect

Electron-scale turbulence is predicted to drive anomalous electron thermal transport. However, experimental study of its relation with transport is still in its early stage. On the National Spherical Tokamak eXperiment (NSTX), electron-scale density ?uctuations are studied with a novel tangen- tial microwave scattering system with high radial resolution of ±2 cm. Here, we report a study of parametric dependence of electron-scale turbulence in NSTX H-mode plasmas. The dependence on density gradient is studied through the observation of a large density gradient variation in the core induced by an ELM event, where we found the ?rst clear experimental evidence of density gradient stabilization of electron-gyro scale turbulence in a fusion plasma. This observation, cou- pled with linear gyro-kinetic calculations, leads to the identi?cation of the observed instability as toroidal Electron Temperature Gradient (ETG) modes. It is observed that longer wavelength ETG modes, k??s < 10 (?s is the ion gyroradius at electron temperature and k? is the wavenumber perpendicular to local equilibrium magnetic ?eld), are most stabilized by density gradient, and the stabilization is accompanied by about a factor of two decrease in electron thermal di?usivity. Comparisons with nonlinear ETG gyrokinetic simulations shows ETG turbulence may be able to explain the experimental electron heat ?ux observed before the ELM event. The collisionality dependence of electron-scale turbulence is also studied by systematically varying plasma current and toroidal ?eld, so that electron gyroradius (?e ), electron beta (?e ) and safety factor (q95 ) are kept approximately constant. More than a factor of two change in electron collisionality, ? ?e, was achieved, and we found that the spectral power of electron-scale turbulence appears to increase as ? ?e is decreased in this collisonality scan. However, both linear and nonlinear simulations show no or weak dependence with the electron-ion collision frequency, ? e/i . Instead, other equilibrium parameters (safety factor, electron density gradient, for example) a?ect ETG linear growth rate and electron thermal transport more than ? e/i does. Furthermore, electron heat ?ux predicted by the simulations is found to have an order-of-magnitude spatial variation in the experimental mea- surement region and is also found to be much smaller than experimental levels except at one radial location we evaluated. The predicted electron heat ?ux is shown to be strongly anti-correlated with density gradient which varies for a factor of three in the measurement region, which is in agreement with the density gradient dependence study reported in this paper.

Ren, Y; Kaye, S M; Mazzucato, E; Bell, R E; Diallo, A; Domier, C W; LeBlanc, B P; Lee, K C; Smith, D R

2012-05-23T23:59:59.000Z

159

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

SciTech Connect

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.

C. L. Smith

2008-08-01T23:59:59.000Z

160

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

SciTech Connect

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.

C. L. Smith

2008-08-01T23:59:59.000Z

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161

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

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.

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

162

Generating optimal states for a homodyne Bell test  

E-Print Network (OSTI)

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.

Sonja Daffer; Peter L. Knight

2005-04-12T23:59:59.000Z

163

Safety of next generation power reactors  

Science Conference Proceedings (OSTI)

This book is organized under the following headings: Future needs of utilities regulators, government, and other energy users, PRA and reliability, LMR concepts, LWR design, Advanced reactor technology, What the industry can deliver: advanced LWRs, High temperature gas-cooled reactors, LMR whole-core experiments, Advanced LWR concepts, LWR technology, Forum: public perceptions, What the industry can deliver: LMRs and HTGRs, Criteria and licensing, LMR modeling, Light water reactor thermal-hydraulics, LMR technology, Working together to revitalize nuclear power, Appendix A, luncheon address, Appendix B, banquet address.

Not Available

1988-01-01T23:59:59.000Z

164

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

E-Print Network (OSTI)

A successful space mission depends on the reliable operation of the spacecraft's electrical power system. For payloads requiring high power levels, various designs of space nuclear power systems (SNPS) are available. Designers have conducted limited spacecraft component reliability analysis and full-scale testing of SNPS is impractical. Therefore, a properly-designed reliability analysis, systematically applied, may provide an effective means for making judgments about the relative reliability of competing SNPSS. This work examines the applicability of probabilistic risk assessment (PRA) techniques for estimating SNPS reliability from design studies. The nuclear electric power industry has used PRA techniques to accurately analyze the reliability of complex systems. However, these PRA techniques for nuclear power plants require modifications for SNPS reliability assessment. This study validates these modified PRA techniques by examining the reliability of the SP-I 00 and the Small Ex-core Heat Pipe Thermionic Reactor (SEHPTR). The present analysis focuses on the SNPS failure to produce nominal electrical power. Typical events threatening the reliability of the SNPS will consist of hardware failures, external events, and command errors or software deficiencies. This work involves the following systematic steps for each SNPS: e System familiarization ³Performance of a "failure modes and effects analysis" to deten-nine how the failures of components might cause a system failure ³Construction of system and component fault trees ³Reliability data estimation³Fault tree quantification (using CAFTA'O and UNCERT'O) 'Me reliability data estimation relies on occurrence probabilities for each component failure mode. Various methods for estimating failure rates from existing reliability databases or from engineering approximations were investigated. This work employs the Monte Carlo sampling technique to associate numerical uncertainty levels with the quantitative reliability estimates produced for each SNPS. The quantitative results estimate the reliability of the systems studied as 0.9494 for the SP-100 and 0.9453 for the SEHPTR. The associated error factor is approximately 2.0, corresponding to the system modeling and reliability data uncertainties. Importance measures and sensitivity analyses indicate that the fuel damage, sensor, electrical component, mechanical component, drive, and power conditioning, control, and distribution subsystem failures can be critical to the system's reliability.

Gutner, Sophie Isabelle

1996-01-01T23:59:59.000Z

165

Adaptive Sampling using Support Vector Machines  

Science Conference Proceedings (OSTI)

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.

D. Mandelli; C. Smith

2012-11-01T23:59:59.000Z

166

Fire Probabilistic Risk Assessment Methods Enhancements  

Science Conference Proceedings (OSTI)

This report documents the interim guidance of the industry and the U.S. Nuclear Regulatory Commission (NRC) on several issues from the National Fire Protection Association (NFPA) Standard 805 Frequently Asked Questions (FAQs) Program arising from use of EPRI 1011989, NUREG/CR-6850 (a joint report of EPRI and the NRC Office of Nuclear Regulatory Research [NRC-RES]) fire PRA methodology for nuclear facilities. The FAQ program was established by the NRC Office of Nuclear Reactor Regulation (NRC/NRR) to supp...

2009-12-23T23:59:59.000Z

167

Review of Quantitative Software Reliability Methods  

Science Conference Proceedings (OSTI)

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.

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

2010-09-17T23:59:59.000Z

168

Review of the Diablo Canyon probabilistic risk assessment  

SciTech Connect

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.

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

169

Improved assessment of aviation hazards to ground facilities using a geographical information system  

SciTech Connect

A computer based system for performing probabilistic safety assessments (PSAs) of aircraft crashes to ground structures is under development. The system called ACRA (aircraft crash risk assessment) employs a GIS (geographical information system) for locating, mapping, and characterizing ground structures; and a multiparameter data base system that supports the analytical PRA (probabilistic risk assessment) model for determining PSAs for aircraft crashes. The Salt Lake City International Airport (SLC) is being employed as the base case for study and application of ACRA and evaluation of the projected safety assessment.

Sandquist, G.M.; Slaughter, D.M. [Utah Univ., Salt Lake City, UT (United States); Kimura, C.Y.

1996-06-03T23:59:59.000Z

170

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

SciTech Connect

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

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

1992-11-01T23:59:59.000Z

171

SAPHIRE 8 Software Project Plan  

SciTech Connect

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.

Curtis L.Smith; Ted S. Wood

2010-03-01T23:59:59.000Z

172

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

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

173

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

SciTech Connect

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.

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

2011-03-13T23:59:59.000Z

174

An approach to estimating radiological risk of offsite release from a design basis earthquake for the Process Experimental Pilot Plant (PREPP)  

SciTech Connect

In compliance with Department of Energy (DOE) Order 6430.1A, a seismic analysis was performed on DOE's Process Experimental Pilot Plant (PREPP), a facility for processing low-level and transuranic (TRU) waste. Because no hazard curves were available for the Idaho National Engineering Laboratory (INEL), DOE guidelines were used to estimate the frequency for the specified design-basis earthquake (DBE). A dynamic structural analysis of the building was performed, using the DBE parameters, followed by a probabilistic risk assessment (PRA). For the PRA, a functional organization of the facility equipment was effected so that top events for a representative event tree model could be determined. Building response spectra (calculated from the structural analysis), in conjunction with generic fragility data, were used to generate fragility curves for the PREPP equipment. Using these curves, failure probabilities for each top event were calculated. These probabilities were integrated into the event tree model, and accident sequences and respective probabilities were calculated through quantification. By combining the sequences failure probabilities with a transport analysis of the estimated airborne source term from a DBE, onsite and offsite consequences were calculated. The results of the comprehensive analysis substantiated the ability of the PREPP facility to withstand a DBE with negligible consequence (i.e., estimated release was within personnel and environmental dose guidelines). 57 refs., 19 figs., 20 tabs.

Lucero, V.; Meale, B.M.; Reny, D.A.; Brown, A.N.

1990-09-01T23:59:59.000Z

175

Preliminary Hazards Analysis Plasma Hearth Process  

SciTech Connect

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.

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

176

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

SciTech Connect

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.

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

177

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

SciTech Connect

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.

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

1994-08-01T23:59:59.000Z

178

A New Class of Risk-Importance Measures to Support Reactor Aging Management and the Prioritization of Materials Degradation Research  

Science Conference Proceedings (OSTI)

As the US fleet of light water reactors ages, the risks of operation might be expected to increase. Although probabilistic risk assessment has proven a critical resource in risk-informed regulatory decision-making, limitations in current methods and models have constrained their prospective value in reactor aging management. These limitations stem principally from the use of static component failure rate models (which do not allow the impact of component aging on failure rates to be represented) and a very limited treatment of passive components (which would be expected to have an increasingly significant risk contribution in an aging system). Yet, a PRA captures a substantial knowledge base that could be of significant value in addressing plant aging. In this paper we will describe a methodology and a new class of risk importance measures that allow the use of an existing PRA model to support the management of plant aging, the prioritization of improvements to non-destructive examination and monitoring techniques, and the establishment of research emphases in materials science. This methodology makes use of data resources generated under the USNRC Proactive Management of Materials Degradation program which addresses the anticipated effects of numerous aging degradation mechanisms on a wide variety of component types.

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

2010-06-07T23:59:59.000Z

179

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

SciTech Connect

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

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

2007-07-01T23:59:59.000Z

180

N reactor individual risk comparison to quantitative nuclear safety goals  

Science Conference Proceedings (OSTI)

A full-scope level III probabilistic risk assessment (PRA) has been completed for N reactor, a US Department of Energy (DOE) production reactor located on the Hanford Reservation in the state of Washington. Sandia National Laboratories (SNL) provided the technical leadership for this work, using the state-of-the-art NUREG-1150 methodology developed for the US Nuclear Regulatory Commission (NRC). The main objectives of this effort were to assess the risks to the public and to the on-site workers posed by the operation of N reactor, to identify changes to the plant that could reduce the overall risk, and to compare those risks to the proposed NRC and DOE quantitative safety goals. This paper presents the methodology adopted by Westinghouse Hanford Company (WHC) and SNL for individual health risk evaluation, its results, and a comparison to the NRC safety objectives and the DOE nuclear safety guidelines. The N reactor results, are also compared with the five NUREG-1150 nuclear plants. Only internal events are compared here because external events are not yet reported in the current draft NUREG-1150. This is the first full-scope level III PRA study with a detailed quantitative safety goal comparison performed for DOE production reactors.

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

1990-01-01T23:59:59.000Z

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181

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

Science Conference Proceedings (OSTI)

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.

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

182

Analysis of Kuosheng Station Blackout Accident Using MELCOR 1.8.4  

Science Conference Proceedings (OSTI)

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

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

2000-11-15T23:59:59.000Z

183

DEGRADATION SUSCEPTIBILITY METRICS AS THE BASES FOR BAYESIAN RELIABILITY MODELS OF AGING PASSIVE COMPONENTS AND LONG-TERM REACTOR RISK  

Science Conference Proceedings (OSTI)

Conventional probabilistic risk assessments (PRAs) are not well-suited to addressing long-term reactor operations. Since passive structures, systems and components are among those for which refurbishment or replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant. Yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA does present challenges. First, service data required to quantify component reliability models are sparse, and this problem is exacerbated by the greater data demands of age-dependent reliability models. A compounding factor is that there can be numerous potential degradation mechanisms associated with the materials, design, and operating environment of a given component. This deepens the data problem since the risk-informed management of materials degradation and component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates the metrics of materials degradation susceptibility being developed under the Nuclear Regulatory Commission's Proactive Management of Materials of Degradation Program with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by the predicted long-term operational risk.

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

2011-07-17T23:59:59.000Z

184

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

SciTech Connect

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.

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

185

Nuclear Material Control and Accountability System Effectiveness Tool (MSET)  

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

186

EPRI/NRC-RES fire human reliability analysis guidelines.  

Science Conference Proceedings (OSTI)

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.

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

187

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

SciTech Connect

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.

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

2013-10-01T23:59:59.000Z

188

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

NLE Websites -- All DOE Office Websites (Extended Search)

Advanced Radiation Medicine Advanced Radiation Medicine Dates Scheduled Register Online March 24-28, 2014 August 18-22, 2014 Fee: $250 Maximum enrollment: 28 30 hours AMA PRA Category 1 Credits(tm) This 4½-day course includes more advanced information for medical practitioners. This program is academically more rigorous than the Radiation Emergency Medicine course and is primarily for Physicians, Clinical Nurse Practitioners and Physician Assistants desiring an advanced level of information on the diagnosis and management of ionizing radiation injuries and illnesses. Advanced topics in the diagnosis and management of radiation-induced injuries and illnesses includes the use of cytokines, stem cell transplants, antimicrobials, wound care and other advanced techniques. Group problem-solving is used to thoroughly orient attendees to the

189

Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Integrated Safety Analysis: Why It Is Appropriate for Fuel Recycling Facilities Executive Summary This paper addresses why the use of an Integrated Safety Analysis ("ISA") is appropriate for fuel recycling facilities 1 which would be licensed under new regulations currently being considered by NRC. The use of the ISA for fuel facilities under Part 70 is described and compared to the use of a Probabilistic Risk Assessment ("PRA") for reactor facilities. A basis is provided for concluding that future recycling facilities - which will possess characteristics similar to today's fuel cycle facilities and distinct from reactors - can best be assessed using established qualitative or semi-quantitative ISA techniques to achieve and demonstrate safety in an effective and efficient manner.

190

Risk Informed Safety Margin Characterization Case Study: Selection of  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Case Study: Selection Case Study: Selection of Electrical Equipment To Be Subjected to Environmental Qualification Risk Informed Safety Margin Characterization Case Study: Selection of Electrical Equipment To Be Subjected to Environmental Qualification Reference 1 discussed key elements of the process for developing a margins-based "safety case" to support safe and efficient operation for an extended period. The present report documents (in Appendix A) a case study, carrying out key steps of the Reference 1 process, using an actual plant Probabilistic Risk Assessment (PRA) model. In general, the margins-based safety case helps the decision-maker manage plant margins most effectively. It tells the plant decision-maker such things as what margin is present (at the plant level, at the functional

191

Wyoming's Economic Future: Planning for Sustained Prosperity  

NLE Websites -- All DOE Office Websites (Extended Search)

Zunsheng Jiao Zunsheng Jiao Senior Geologist WSGS Future Work * Refine the geological framework required for 3-D rock fluid modeling of the Rock Springs Uplift (RSU). * Construct a 3-D numerical model of CO 2 injection into the RSU. * Build a Performance Assessment (PA) model that includes uncertainty and that can be utilized to construct a Probabilistic Risk Analysis (PRA) for CO 2 sequestration at the RSU. A SYSTEM MODEL FOR GEOLOGIC SEQUESTRATION OF CARBON DIOXIDE CO2_PENS, Los Alamos/Goldsim Rock Springs Uplift: an outstanding geological CO 2 sequestration site in southwestern Wyoming * Thick saline aquifer sequence overlain by thick sealing lithologies. * Doubly-plunging anticline characterized by more than 10,000 ft of closed structural relief. * Huge area (50 x 35 mile).

192

ORISE: Radiation Emergency Medicine - Continuing Medical Education Course  

NLE Websites -- All DOE Office Websites (Extended Search)

Radiation Emergency Medicine Radiation Emergency Medicine Dates Scheduled Register Online February 4-7, 2014 March 18-21, 2014 April 29-May 2, 2014 June 3-6, 2014 August 12-15, 2014 Fee: $175 Maximum enrollment: 24 24.5 hours AMA PRA Category 1 Credits(tm) This 3½-day course is intended for physicians, nurses, nurse practitioners and physician assistants who may be called upon to provide emergency medical care following a radiological or nuclear incident. Priority registration will be given to these groups of professionals. This course may also be relevant for paramedic instructors but is generally not intended for pre-hospital responders. The course emphasizes the practical aspects of initial hospital management of irradiated and/or contaminated patients through lectures and hands-on practical exercises.

193

Notices  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

5 Federal Register 5 Federal Register / Vol. 78, No. 93 / Tuesday, May 14, 2013 / Notices * Mail/Hand Delivery/Courier: Consumer Financial Protection Bureau (Attention: PRA Office), 1700 G Street NW., Washington, DC 20552. Please note that comments submitted by fax or email and those submitted after the comment period will not be accepted. In general, all comments received will be posted without change to regulations.gov, including any personal information provided. Sensitive personal information, such as account numbers or social security numbers, should not be included. FOR FURTHER INFORMATION CONTACT: Documentation prepared in support of this information collection request is available at www.regulations.gov. Requests for additional information should be directed to the Consumer

194

Notices  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

200 Federal Register 200 Federal Register / Vol. 79, No. 12 / Friday, January 17, 2014 / Notices be collecting are available in the public docket for this ICR. The docket can be viewed online at www.regulations.gov or in person at the EPA Docket Center, WJC West, Room 3334, 1301 Constitution Ave. NW., Washington, DC. The telephone number for the Docket Center is 202-566-1744. For additional information about EPA's public docket, visit http://www.epa.gov/ dockets. Pursuant to section 3506(c)(2)(A) of the PRA, EPA specifically solicits comments and information to enable it to: (i) Evaluate whether the proposed collection of information is necessary for the proper performance of the functions of the Agency, including whether the information will have practical utility; (ii) evaluate the

195

Frequently Asked Questions | U.S. DOE Office of Science (SC)  

Office of Science (SC) Website

Frequently Frequently Asked Questions Science Undergraduate Laboratory Internships (SULI) SULI Home Eligibility Benefits Participant Obligations How to Apply Key Dates Frequently Asked Questions Contact WDTS Home Frequently Asked Questions Print Text Size: A A A RSS Feeds FeedbackShare Page TABLE OF CONTENTS General Eligibility Applications Selection Participation Sponsor GENERAL When is the application deadline? The application deadline for the 2014 Summer SULI Term is January 10, 2014 5:00 PM ET. Who administers this program for the Department of Energy? The DOE Office of Workforce Development for Teachers and Scientists (WDTS) manages this program in collaboration with the DOE National Laboratories who host the student participants. Back to Top Back to Top ELIGIBILITY Can I apply if I don't currently have legal permanent resident alien (PRA)

196

Frequently Asked Questions | U.S. DOE Office of Science (SC)  

Office of Science (SC) Website

Frequently Frequently Asked Questions Community College Internships (CCI) CCI Home Eligibility Benefits Participant Obligations How to Apply Key Dates Frequently Asked Questions Contact WDTS Home Frequently Asked Questions Print Text Size: A A A RSS Feeds FeedbackShare Page TABLE OF CONTENTS General Eligibility Applications Selection Participation Sponsor GENERAL When is the application deadline? The application deadline for the 2014 Summer CCI Term is January 10, 2014 5:00 PM ET. Who administers the CCI program for the Department of Energy? The DOE Office of Workforce Development for Teachers and Scientists manages this program in collaboration with the DOE National Laboratories who host the student participants. Back to Top Back to Top ELIGIBILITY Can I apply if I don't currently have legal permanent resident alien (PRA)

197

Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis |  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Nuclear Energy Institute (NEI) Attachment, Integrated Safety Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis Nuclear Energy Institute (NEI) Attachment, Integrated Safety Analysis This paper addresses why the use of an Integrated Safety Analysis ("ISA") is appropriate for fuel recycling facilities1 which would be licensed under new regulations currently being considered by NRC. The use of the ISA for fuel facilities under Part 70 is described and compared to the use of a Probabilistic Risk Assessment ("PRA") for reactor facilities. A basis is provided for concluding that future recycling facilities - which will possess characteristics similar to today's fuel cycle facilities and distinct from reactors - can best be assessed using established qualitative or semi-quantitative ISA techniques to achieve and

198

Physics-Based Stress Corrosion Cracking Component Reliability Model cast in  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Physics-Based Stress Corrosion Cracking Component Reliability Model Physics-Based Stress Corrosion Cracking Component Reliability Model cast in an R7-Compatible Cumulative Damage Framework Physics-Based Stress Corrosion Cracking Component Reliability Model cast in an R7-Compatible Cumulative Damage Framework The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). The methodology emerging from the RISMC pathway is not a conventional probabilistic risk assessment (PRA)-based one; rather, it relies on a reactor systems simulation framework in which

199

DOE F 4200.33.cdr  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

3 3 (07-05) U.S. Department of Energy Note: ** We Hereby Certify That Funds Cited Are Proper For This Procurement And In Compliance With Applicable Appropriations Acts and Fiscal Law. Printed with soy ink on recycled paper Procurement Request-Authorization 1. Awarding Office Fund Year Alottee Reporting Entity SGL Object Class Program Project WFO Local Use 26. Dollar Amount 27. Program Budget Official's Signature** 3. PRA Number Formerly PR-799A (Previous editions are obsolete) 2. Initiating Office 4. Change/Correction in Process? Yes No INITIAL PADS DATA ENTRY INFORMATION 5. Description of Work/Purpose of Assistance 6. Awardee Name 6a. Division Yes No Has List of Sources Been Attached? 7. Address 8. Government Share 9. Awardee Share

200

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

31 - 14940 of 26,764 results. 31 - 14940 of 26,764 results. Download PSH-12-0123- In the Matter of Personnel Security Hearing On February 15, 2013, an OHA Hearing Officer issued a decision in which he determined that the DOE should not restore an individual's access authorization. As security concerns under 10 CFR Part... http://energy.gov/oha/downloads/psh-12-0123-matter-personnel-security-hearing Download Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11 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... http://energy.gov/hss/downloads/comparison-integrated-safety-analysis-isa-and-probabilistic-risk

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
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201

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

61 - 24270 of 28,905 results. 61 - 24270 of 28,905 results. Page Information Management and Supporting Documentation 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... http://energy.gov/cio/information-management-and-supporting-documentation Page About the Library The Law Library is located on the 6th floor of Forrestal between the A and B corridors at 6A-156 or 6B-157. The doors to the Law Library remain open as long as the Forrestal Building is accessible... http://energy.gov/gc/about-library Article Secretary Chu honors America's Nuclear Security Workers Remarks highlight past service, current accomplishments http://energy.gov/articles/secretary-chu-honors-americas-nuclear-security-workers

202

Physics-Based Stress Corrosion Cracking Component Reliability Model cast in  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Physics-Based Stress Corrosion Cracking Component Reliability Model Physics-Based Stress Corrosion Cracking Component Reliability Model cast in an R7-Compatible Cumulative Damage Framework Physics-Based Stress Corrosion Cracking Component Reliability Model cast in an R7-Compatible Cumulative Damage Framework The Risk-Informed Safety Margin Characterization (RISMC) pathway is a set of activities defined under the U.S. Department of Energy Light Water Reactor Sustainability Program. The overarching objective of RISMC is to support plant life-extension decision-making by providing a state-of-knowledge characterization of safety margins in key systems, structures, and components (SSCs). The methodology emerging from the RISMC pathway is not a conventional probabilistic risk assessment (PRA)-based one; rather, it relies on a reactor systems simulation framework in which

203

5th International REAC/TS Symposium: The Medical Basis for Radiation  

NLE Websites -- All DOE Office Websites (Extended Search)

Privacy/Security Statement Privacy/Security Statement 5th International REAC/TS Symposium: The Medical Basis for Radiation Accident Preparedness Skip site navigation and move to main content of page. Home Schedule Speakers Registration Directions and Acommodations Contact 5th International REAC/TS Symposium: The Medical Basis for Radiation Accident Preparedness Sept. 27-29, 2011 Hilton Miami Downtown Miami, Florida United States Introduction This symposium brings together international experts to discuss the advances in the diagnosis and management of radiation emergencies and illnesses. The Oak Ridge Institute for Science and Education (ORISE) designates this live activity for a maximum of 19.25 AMA PRA Category 1 Credit(s)(tm). Physicians should claim only the credit commensurate with the extent of

204

Page not found | Department of Energy  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

401 - 2410 of 26,777 results. 401 - 2410 of 26,777 results. Download Solar Background Document 1 A timeline outlining the Energy Department's extensive review of the Solyndra Solar loan guarantee application from 2006 to 2009. http://energy.gov/downloads/solar-background-document-1 Download TEC Working Group Topic Groups Rail Key Documents Intermodal Subgroup Intermodal Subgroup http://energy.gov/em/downloads/tec-working-group-topic-groups-rail-key-documents-intermodal-subgroup Page Information Management and Supporting Documentation 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... http://energy.gov/cio/information-management-and-supporting-documentation

205

Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight  

Energy.gov (U.S. Department of Energy (DOE)) Indexed Site

Safety Culture in the US Nuclear Regulatory Commission's Reactor Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process September 19, 2012 Presenter: Undine Shoop, Chief, Health Physics and Human Performance Branch, Office of Nuclear Reactor Regulation, U.S. Nuclear Regulatory Commission Topics covered: Purpose of the Reactor Oversight Process (ROP) ROP Framework Safety Culture within the ROP Safety Culture Assessments Safety Culture in the US Nuclear Regulatory Commission's Reactor Oversight Process More Documents & Publications A Commissioner's Perspective on USNRC Actions in Response to the Fukushima Nuclear Accident Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk Assessment (PRA) for Fuel Cycle Facilities, 2/17/11

206

NETL: Methane Hydrates - Barrow Gas Fields - North Slope Borough, Alaska  

NLE Websites -- All DOE Office Websites (Extended Search)

Phase 2- Drilling and Production Testing the Methane Hydrate Resource Potential associated with the Barrow Gas Fields Last Reviewed 04/06/2010 Phase 2- Drilling and Production Testing the Methane Hydrate Resource Potential associated with the Barrow Gas Fields Last Reviewed 04/06/2010 DE-FC26-06NT42962 Goal The goal of this project is to evaluate, design, drill, log, core and production test methane hydrate resources in the Barrow Gas Fields near Barrow, Alaska to determine its impact on future free gas production and its viability as an energy source. Photo of Barrow welcome sign Performers North Slope Borough, Barrow, Alaska 99723 Petrotechnical Resources Alaska (PRA), Fairbanks, AK 99775 University of Alaska Fairbanks, Fairbanks, AK 99775 Background Phase 1 of the Barrow Gas Fields Hydrate Study provided very strong evidence for the existence of hydrates updip of the East Barrow and Walakpa Gas Fields. Full-field history matched reservoir modeling supported the

207

Regulatory cross-cutting topics for fuel cycle facilities.  

Science Conference Proceedings (OSTI)

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.)

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

2013-10-01T23:59:59.000Z

208

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

SciTech Connect

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

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

1989-08-01T23:59:59.000Z

209

Integrated systems analysis of the PIUS reactor  

SciTech Connect

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.

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

1993-11-01T23:59:59.000Z

210

Les bKa'brgyad - Sources canoniques et tradition de Nyangral Nyima 'od zer  

E-Print Network (OSTI)

la Pensée, lié à la famille du bouddha Ak?obhya (Thugs Mi bskyod pa'i rigs) nommé dPal Heruka snying rje rol pa'i rgyud (?r?-heruka-karu??kr??ita-tantra), établi par l'?c?rya H??- kara12 , venu de l'Inde de l'est, près de Zahor (Est du Bengale) (r... mi tra, Pra chen ha ti, Rum bu ghu ya dhe ba, Dha na sang tri, Shan ting ghar ba. 21 'Ju mi pham (s.d., vol. 21 : 15) mentionne également cette neuvième cassette, dont les autres gter ston qui ont découvert des cycles de bKa' brgyad ne parlent pas...

Sampel, Tenzin

2008-01-01T23:59:59.000Z

211

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

Science Conference Proceedings (OSTI)

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.

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

1995-01-01T23:59:59.000Z

212

Evaluation of potential severe accidents during low power and shutdown operations at Surry, Unit 1: Evaluation of severe accident risk during mid-loop operations. Volume 6, Part 2: Appendices  

SciTech Connect

The objectives 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. In phase 2, mid-loop operation was selected as the plant configuration to be analyzed. Volume 1 summarizes the results of the study. The scope of the level-1 study includes plant damage state analyses, and uncertainty analysis. The internal event analysis is documented in Volume 2. The internal fire and internal flood analysis are documented in Volumes 3 and 4, respectively. A separate study on seismic analysis, documented in Volume 5, was performed for the NRC by Future Resources Associated, Inc. A phased approach was used in the level 2/3 PRA program, however both phases addressed the risk from only mid-loop operation. The first phase of the level 2/3 PRA was initiated in late 1991 and consisted of an Abridged Risk Study. This study was completed in May 1992 and was focused on accident progression and consequences, conditional on core damage. Phase 2 is a more detailed study in which an evaluation of risk during mid-loop operation was performed. The results of the phase 2 level 2/3 study are the subject of this volume of NUREG/CR-6144, Volume 6. This report, Volume 6, Part 2, consists of five appendices containing supporting information for: the PDS (plant damage state) analysis; the accident progression analysis; the source term analysis; the consequence analysis; and the Melcor analysis. 73 figs., 21 tabs.

Jo, J.; Lin, C.C.; Neymotin, L.; Mubayi, V. [Brookhaven National Lab., Upton, NY (United States)

1995-05-01T23:59:59.000Z

213

RELIABILITY MODELS OF AGING PASSIVE COMPONENTS INFORMED BY MATERIALS DEGRADATION METRICS TO SUPPORT LONG-TERM REACTOR OPERATIONS  

Science Conference Proceedings (OSTI)

Paper describes a methodology for the synthesis of nuclear power plant service data with expert-elicited materials degradation information to estimate the future failure rates of passive components. This method should be an important resource to long-term plant operations and reactor life extension. Conventional probabilistic risk assessments (PRAs) are not well suited to addressing long-term reactor operations. Since passive structures and components are among those for which replacement can be least practical, they might be expected to contribute increasingly to risk in an aging plant; yet, passives receive limited treatment in PRAs. Furthermore, PRAs produce only snapshots of risk based on the assumption of time-independent component failure rates. This assumption is unlikely to be valid in aging systems. The treatment of aging passive components in PRA presents challenges. Service data to quantify component reliability models are sparse, and this is exacerbated by the greater data demands of age-dependent reliability models. Another factor is that there can be numerous potential degradation mechanisms associated with the materials and operating environment of a given component. This deepens the data problem since risk-informed management of component aging will demand an understanding of the long-term risk significance of individual degradation mechanisms. In this paper we describe a Bayesian methodology that integrates metrics of materials degradation susceptibility with available plant service data to estimate age-dependent passive component reliabilities. Integration of these models into conventional PRA will provide a basis for materials degradation management informed by predicted long-term operational risk.

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

2012-05-01T23:59:59.000Z

214

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  

Science Conference Proceedings (OSTI)

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.

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

215

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

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.

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

216

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

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.

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

217

Conversion of Questionnaire Data  

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

218

RECENT DEVELOPMENTS OF THE IRIS PROJECT OF INTEREST FOR LATIN AMERICA  

Science Conference Proceedings (OSTI)

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

Carelli, M.D.; Petrovic, B.

2004-10-03T23:59:59.000Z

219

An experimental study of assessment of weld quality on fatigue reliability analysis of a nuclear pressure vessel  

SciTech Connect

The steam generator in a PWR primary coolant system is one of the pieces of equipment made in China for the Qinshan nuclear power plant, Zhejiang. It is a crucial unit belonging to the category of nuclear pressure vessel. The purpose of this research work is to carry out an examination of the weld quality of the steam generator under fatigue loading and to assess its reliability by using experimental results of a fatigue test of the nuclear pressure vessel steel S-271 (Chinese Standard) and of qualified tests of welded seams of a simulated prototype of the bottom closure head of the steam generator. A guarantee of weld quality is proposed as the quality assurance of safety for the China National Nuclear Safety Supervision Bureau. The results of the reliability analysis reported in this work can be taken as supplementary material for a Probabilistic Risk Assessment (PRA) of the Qinshan nuclear power plant. According to the requirement of Provision 2-1500 CYCLIC TESTING, ASME Boiler and Pressure Vessel Code, Section 3, Rules for Construction of Nuclear Power Plant Components, a simulated prototype of the bottom closure head of the steam generator was made in this work for the qualified tests. Qualified tests with small sample size present a problem which is difficult to solve in reliability analysis, and are therefore of interest. Here, the authors offer proposals attempting to solve this problem.

Dai, Shuho (Nanjing Inst. of Chemical Technology, Jiangsu (China). Dept. of Mechanical Engineering)

1993-11-01T23:59:59.000Z

220

Randomized benchmarking of single and multi-qubit control in liquid-state NMR quantum information processing  

E-Print Network (OSTI)

Being able to quantify the level of coherent control in a proposed device implementing a quantum information processor (QIP) is an important task for both comparing different devices and assessing a device's prospects with regards to achieving fault-tolerant quantum control. We implement in a liquid-state nuclear magnetic resonance QIP the randomized benchmarking protocol presented by Knill et al (PRA 77: 012307 (2008)). We report an error per randomized $\\frac{\\pi}{2}$ pulse of $1.3 \\pm 0.1 \\times 10^{-4}$ with a single qubit QIP and show an experimentally relevant error model where the randomized benchmarking gives a signature fidelity decay which is not possible to interpret as a single error per gate. We explore and experimentally investigate multi-qubit extensions of this protocol and report an average error rate for one and two qubit gates of $4.7 \\pm 0.3 \\times 10^{-3}$ for a three qubit QIP. We estimate that these error rates are still not decoherence limited and thus can be improved with modifications to the control hardware and software.

C. A. Ryan; M. Laforest; R. Laflamme

2008-08-28T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


221

E/EIA  

U.S. Energy Information Administration (EIA) Indexed Site

E/EIA E/EIA -0278 U.S. Depa rtme nt of Energ y Energ y Inform ation Admi nistra tion Assis tant Admi nistra tor for Progr am Deve lopme nt Office of the Cons umpt ion Data Syste m June 1981 01377 9 = 4530 : FED Non res ide ntia l Bui ldin gs u/w & Ene rgy Con sum ptio n Sur vey : Fu el Ch ara cte ris tic s an d Co ns erv ati on Pra cti ces Prepared by: Lynn D. Patinkin, Phillip Windell, Dwight: K. French, Leigh Carleton, Lynda T. Carlson, Kenneth A. Vagts, Leslie Whitaker, Tom Woteki, Wilbert Laird, and Laura Wong. IMPORTANT NOTICE As required by government regulation, EIA will conduct the annual review of our mailing list during the next several weeks. If you are on the mailing list, you will soon receive a post card listing your name and address as they appear on our files. If you wish to continue to receive our publications, you must mail

222

Electricity 2011  

U.S. Energy Information Administration (EIA) Indexed Site

Electricity > Soliciting comments on EIA-111 Electricity > Soliciting comments on EIA-111 EIA announces the proposal of Form EIA-111, Quarterly Electricity Imports and Exports Report Released: August 15, 2011 Background On August 11, 2011, a Federal Register Notice was published soliciting comments for the new EIA-111 survey form. The EIA-111, Quarterly Electricity Imports and Exports Report will replace the OE-781R, Monthly Electricity Imports and Exports Report. The OE-781R has been suspended and will be terminated upon the approval of the EIA-111. The OE-781R administered from July 2010 through May 2011, proved complex and confusing for the repondents. As a result, the EIA-111 was developed to more effectively and efficiently collect more accurate and meaningful data. The Paperwork Reduction Act (PRA) of 1995 requires that each Federal agency obtains approval from the Office of Management and Budget (OMB) before undertaking to collect information from ten or more persons, or continuing a collection for which the OMB approval and the OMB control number are about to expire. The approval process, which is popularly known as the "OMB clearance process," is extensive. It requires two Federal Register notices and a detailed application ("supporting statement") to OMB. The first Federal Register Notice was published on August 11, 2011. EIA is prepared to address the comments submitted by each individual.

223

Integrated safety assessment of an oxygen reduction project at Connecticut Yankee Atomic Power's Haddam Neck plant  

SciTech Connect

Connecticut Yankee Atomic Power Company (CYAPCo) has implemented an Integrated Safety Assessment Program (ISAP) for the integrated evaluation and prioritization of plant-specific licensing issues, regulatory policy issues, and plant improvement projects. As part of the ISAP process, probabilistic risk assessment (PRA) is utilized to evaluate the net safety impact of plant modification projects. On a few occasions, implementation of this approach has resulted in the identification of projects with negative safety impacts that could not be quantified via the normal design review and 10CFR50.59 safety evaluation process. An example is a plant modification that was proposed to reduce the oxygen in the Haddam Neck plant's demineralized water storage tank (DWST). The project involved the design and installation of a nitrogen blanketing system on the DWST. The purpose of the project was to reduce the oxygen content on the secondary side, consistent with recommendations from the Electric Power Research Institute Steam Generator Owners Group. Oxygen is one of the contributors to the corrosion process in systems in contact with the feedwater and can cause damage to associated components if not controlled.

Aubrey, J.E.

1987-01-01T23:59:59.000Z

224

Risk Insights Gained from Fire Incidents  

SciTech Connect

There now exist close to 20 years of history in the application of Probabilistic Risk Assessment (PRA) for the analysis of fire risk at nuclear power plants. The current methods are based on various assumptions regarding fire phenomena, the impact of fire on equipment and operator response, and the overall progression of a fire event from initiation through final resolution. Over this same time period, a number of significant fire incidents have occurred at nuclear power plants around the world. Insights gained from US experience have been used in US studies as the statistical basis for establishing fire initiation frequencies both as a function of the plant area and the initiating fire source.To a lesser extent, the fire experience has also been used to assess the general severity and duration of fires. However, aside from these statistical analyses, the incidents have rarely been scrutinized in detail to verify the underlying assumptions of fire PRAs. This paper discusses an effort, under which a set of fire incidents are being reviewed in order to gain insights directly relevant to the methods, data, and assumptions that form the basis for current fire PRAs. The paper focuses on the objectives of the effort, the specific fire events being reviews methodology, and anticipated follow-on activities.

Kazarians, Mardy; Nowlen, Steven P.

1999-06-10T23:59:59.000Z

225

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

Science Conference Proceedings (OSTI)

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

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

2010-06-01T23:59:59.000Z

226

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

Science Conference Proceedings (OSTI)

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

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

2011-10-01T23:59:59.000Z

227

Development of Simplified Probabilistic Risk Assessment Model for Seismic Initiating Event  

Science Conference Proceedings (OSTI)

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).

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

2012-06-01T23:59:59.000Z

228

Development and experimental validation of computational methods to simulate abnormal thermal and structural environments  

Science Conference Proceedings (OSTI)

Over the past 40 years, Sandia National Laboratories (SNL) has been actively engaged in research to improve the ability to accurately predict the response of engineered systems to abnormal thermal and structural environments. These engineered systems contain very hazardous materials. Assessing the degree of safety/risk afforded the public and environment by these engineered systems, therefore, is of upmost importance. The ability to accurately predict the response of these systems to accidents (to abnormal environments) is required to assess the degree of safety. Before the effect of the abnormal environment on these systems can be determined, it is necessary to ascertain the nature of the environment. Ascertaining the nature of the environment, in turn, requires the ability to physically characterize and numerically simulate the abnormal environment. Historically, SNL has demonstrated the level of safety provided by these engineered systems by either of two approaches: (1) a purely regulatory approach, or (2) by a Probabilistic Risk Assessment (PRA). This paper will address the latter of the two approaches.

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

1993-10-01T23:59:59.000Z

229

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

Science Conference Proceedings (OSTI)

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

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

1995-08-01T23:59:59.000Z

230

SPAR Model Structural Efficiencies  

SciTech Connect

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

John Schroeder; Dan Henry

2013-04-01T23:59:59.000Z

231

Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) Technical Reference Manual  

Science Conference Proceedings (OSTI)

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. Herein information is provided on the principles used in the construction and operation of Version 6.0 and 7.0 of the SAPHIRE system. This report summarizes the fundamental mathematical concepts of sets and logic, fault trees, and probability. This volume then describes the algorithms used to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the probability of the top event from the minimal cut sets, and the formulas for probabilities that apply for various assumptions concerning reparability and mission time. It defines the measures of basic event importance that SAPHIRE can calculate. This volume gives an overview of uncertainty analysis using simple Monte Carlo sampling or Latin Hypercube sampling, and states the algorithms used by this program to generate random basic event probabilities from various distributions. Also covered are enhance capabilities such as seismic analysis, cut set "recovery," end state manipulation, and use of "compound events."

C. L. Smith; W. J. Galyean; S. T. Beck

2006-07-01T23:59:59.000Z

232

Systems Analysis Programs for Hands-On Integrated Reliability Evaluations (SAPHIRE) Technical Reference  

Science Conference Proceedings (OSTI)

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. Herein information is provided on the principles used in the construction and operation of Version 6.0 and 7.0 of the SAPHIRE system. This report summarizes the fundamental mathematical concepts of sets and logic, fault trees, and probability. This volume then describes the algorithms used to construct a fault tree and to obtain the minimal cut sets. It gives the formulas used to obtain the probability of the top event from the minimal cut sets, and the formulas for probabilities that apply for various assumptions concerning reparability and mission time. It defines the measures of basic event importance that SAPHIRE can calculate. This volume gives an overview of uncertainty analysis using simple Monte Carlo sampling or Latin Hypercube sampling, and states the algorithms used by this program to generate random basic event probabilities from various distributions. Also covered are enhance capabilities such as seismic analysis, cut set "recovery," end state manipulation, and use of "compound events."

C. L. Smith; W. J. Galyean; S. T. Beck

2008-08-01T23:59:59.000Z

233

Regulatory cross-cutting topics for fuel cycle facilities.  

SciTech Connect

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.)

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

2013-10-01T23:59:59.000Z

234

The Multi-Configurational Hartree-Fock close-coupling ansatz: application to Argon photoionization cross section and delays  

E-Print Network (OSTI)

We present a robust, ab initio method for addressing atom-light interactions and apply it to photoionization of argon. We use a close-coupling ansatz constructed on a multi-configurational Hartree-Fock description of localized states and B-spline expansions of the electron radial wave functions. In this implementation, the general many-electron problem can be tackled thanks to the use of the ATSP2K libraries [CPC 176 (2007) 559]. In the present contribution, we combine this method with exterior complex scaling, thereby allowing for the computation of the complex partial amplitudes that encode the whole dynamics of the photoionization process. The method is validated on the 3s3p6np series of resonances converging to the 3s extraction. Then, it is used for computing the energy dependent differential atomic delay between 3p and 3s photoemission, and agreement is found with the measurements of Gu\\'enot et al. [PRA 85 (2012) 053424]. The effect of the presence of resonances in the one-photon spectrum on photoioniz...

Carette, T; Argenti, L; Lindroth, E

2013-01-01T23:59:59.000Z

235

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

SciTech Connect

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

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

1983-08-01T23:59:59.000Z

236

The application of modern safety criteria to restarting and operating the USDOE K-Reactor  

SciTech Connect

The United States Department of Energy's (USDOE's) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992.

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-01-01T23:59:59.000Z

237

The application of modern safety criteria to restarting and operating the USDOE K-Reactor  

SciTech Connect

The United States Department of Energy`s (USDOE`s) K-reactor, a defense production reactor located at the Savannah River Site in Aiken, South Carolina, was shut down in the summer of 1988 for safety upgrades to bring it into conformance with modern safety standards prior to restart. Over the course of the succeeding four years, all aspects of the 35-year old reactor, including hardware, operations, and analysis, were upgraded to ensure that the reactor could operate safely according to standards similar to those applied to modern nuclear reactors. This paper describes the decision making processes by which issues were identified, priorities assigned, and analysis improved to enhance reactor safety. Special emphasis is given to the probabilistic risk assessment (PRA) decision making processes used to quantify the risks and consequences of operating the K-reactor, the analytical hierarchy process (AHP) used to identify key phenomena, and modifications made to the RELAP5 computer code to make it applicable to K-reactor analysis. The success of the project was demonstrated when the K-reactor was restarted in the summer of 1992.

Dimenna, R.A.; Taylor, G.A.; Brandyberry, M.D.

1993-06-01T23:59:59.000Z

238

Mitigation of direct containment heating and hydrogen combustion events in ice condenser plants  

DOE Green Energy (OSTI)

Using Sequoyah as a representative plant, calculations have been performed with a developmental version of the CONTAIN computer code to assess the effectiveness of various possible improvements to ice condenser containments in mitigating severe accident scenarios involving direct containment heating (DCH) and/or hydrogen combustion. Mitigation strategies considered included backup power for igniters and/or air return fans, augmented igniter systems, containment venting, containment inerting, subatmospheric containment operation, reduced ice condenser bypass, and primary system depressurization. Various combinations of these improvements were also considered. Only inerting the containment or primary system depressurization combined with backup power supplies for the igniter systems resulted in large decreases in the peak pressures calculated to result from DCH events. Potential hydrogen detonation threats were also assessed; providing backup power for both the igniter systems and the air return fans would significantly reduce the potential for detonations but might not totally eliminate it. Sensitivity studies using the NUREG-1150 PRA methodology indicated that primary system depressurization combined with backup power for both igniters and fans could reduce the contribution to the mean risk potential of the class of events considered by about a factor of three. 7 refs., 6 figs., 6 tabs.

Williams, D.C.; Gregory, J.J. (Sandia National Labs., Albuquerque, NM (USA))

1990-10-01T23:59:59.000Z

239

HTGR: an assessment of safety and investment risk  

SciTech Connect

Improvements in the present LWR designs which do not change the basic features that raise safety and economic concerns would not be expected to change public perception or be sufficient to encourage new nuclear investments by the utilities. The HTGR offers an attractive alternative, an alternative with proven operating experience and safety characteristics. This is demonstrated in this paper by an assessment of HTGR inherent safety features and by examining Fort St. Vrain operating experience data, which establish the HTGR as a forgiving design with respect to potential accidents. A further quantification of HTGR safety is made from a probabilistic risk assessment (PRA) of the 2240 MWt High-Temperature Gas-Cooled Reactor-Steam Cycle/Cogeneration (HTGR-SC/C) reference plant. These results are compared to NRC risk goals and achieved LWR safety. Finally, a proposed small HTGR reactor design is discussed. Although the relative costs and marketability of such a small reactor have yet to be determined, several additional passive safety characteristics of the small HTGR make such a plant essentially benign.

Fisher, C.; Fortescue, P.; Goodjohn, A.J.; Olsen, B.E.; Silady, F.A.

1984-11-01T23:59:59.000Z

240

Risk assessment handbook  

SciTech Connect

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

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

1990-09-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
While these samples are representative of the content of NLEBeta,
they are not comprehensive nor are they the most current set.
We encourage you to perform a real-time search of NLEBeta
to obtain the most current and comprehensive results.


241

Risk assessment handbook  

SciTech Connect

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

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

1990-09-01T23:59:59.000Z

242

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

SciTech Connect

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

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

1989-08-01T23:59:59.000Z

243

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

SciTech Connect

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

CHENG,L.Y.; LUDEWIG, H.

2007-06-01T23:59:59.000Z

244

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

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.

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

1995-03-01T23:59:59.000Z

245

Human Reliability Analysis for Small Modular Reactors  

Science Conference Proceedings (OSTI)

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.

Ronald L. Boring; David I. Gertman

2012-06-01T23:59:59.000Z

246

Lessons Learned From Implementation of Westinghouse Owners Group Risk-Informed Inservice Inspection Methodology for Piping  

SciTech Connect

Risk-informed inservice inspection (ISI) programs have been in use for over seven years as an alternative to current regulatory requirements in the development and implementation of ISI programs for nuclear plant piping systems. Programs using the Westinghouse Owners Group (WOG) (now known as the Pressurized Water Reactor Owners Group - PWROG) risk-informed ISI methodology have been developed and implemented within the U.S. and several other countries. Additionally, many plants have conducted or are in the process of conducting updates to their risk-informed ISI programs. In the development and implementation of these risk-informed ISI programs and the associated updates to those programs, the following important lessons learned have been identified and are addressed. Concepts such as 'loss of inventory', which are typically not modeled in a plant's probabilistic risk assessment (PRA) model for all systems. The importance of considering operator actions in the identification of consequences associated with a piping failure and the categorization of segments as high safety significant (HSS) or low safety significant (LSS). The impact that the above considerations have had on the large early release frequency (LERF) and categorization of segments as HSS or LSS. The importance of automation. Making the update process more efficient to reduce costs associated with maintaining the risk-informed ISI program. The insights gained are associated with many of the steps in the risk-informed ISI process including: development of the consequences associated with piping failures, categorization of segments, structural element selection and program updates. Many of these lessons learned have impacted the results of the risk-informed ISI programs and have impacted the updates to those programs. This paper summarizes the lessons learned and insights gained from the application of the WOG risk-informed ISI methodology in the U.S., Europe and Asia. (authors)

Stevenson, Paul R.; Haessler, Richard L. [Westinghouse Electric Company, LLC (United States); McNeill, Alex [Dominion Energy, Innsbrook Technical Center (United States); Pyne, Mark A. [Duke Energy (United States); West, Raymond A. [Dominion Nuclear Connecticut, Inc. - Dominion Generation (United States)

2006-07-01T23:59:59.000Z

247

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

E-Print Network (OSTI)

Due to the scarcity of Human Reliability Analysis (HRA) data, one of the key elements of any HRA analysis is use of engineering judgment. The Electric Power Research Institute (EPRI) HRA Calculator guides the user through the steps of any HRA analysis and allows the user to choose among analytical HRA methods. It applies Accident Sequence Evaluation Program (ASEP), Technique for Human Error Rate Prediction (THERP), the HCR/ORE Correlation, and the Caused Based Decision Tree Method (CBDTM). This program is intended to produce consistent results among different analysts provided that the initial information is similar. Even with this analytical approach, an HRA analyst must still render several judgments. The objective of this study was to evaluate the use of engineering judgment applied to the quantification of post-initiator actions using the HRA Calculator. The Comanche Peak Steam Electric Station (CPSES) Level 1 Probabilistic Risk Assessment (PRA) HRA was used as a database for examples and numerical comparison. Engineering judgments were evaluated in the following ways: 1) Survey of HRA experts. Two surveys were completed, and the participants provided a range of different perspectives on how they individually apply engineering judgment. 2) Numerical comparison among the three methods. 3) Review of CPSES HRA and identification of judgments and the effects on the overall results of the database. The results of this study identified thirteen areas in which an HRA analyst must interpret and render judgments on how to quantify a Human Error Probability (HEP) and recommendations are provided on how current industry practitioners render these same judgments. The areas are: identification and definition of actions to be modeled, identification and definition of actions to be modeled, definition of critical actions, definition of cognitive portion of the action, choice of methodology, stress level, rule-, skill- or knowledge-based designation, timing information, training, procedures, human interactions with hardware, recoveries and dependencies within an action, and review of final HEP.

Kohlhepp, Katherine D.

2005-12-01T23:59:59.000Z

248

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

SciTech Connect

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.

NONE

1997-12-01T23:59:59.000Z

249

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

SciTech Connect

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.

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

2008-10-01T23:59:59.000Z

250

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

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.

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

251

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

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.

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

252

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

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.

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

1995-05-01T23:59:59.000Z

253

System Effectiveness  

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

254

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

SciTech Connect

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.

Steve McRae; Thomas Walsh; Michael Dunn; Michael Cook

2010-02-22T23:59:59.000Z

255

Summary  

SciTech Connect

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.

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

2011-01-01T23:59:59.000Z

256

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

SciTech Connect

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.

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

2010-06-30T23:59:59.000Z