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1

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

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

2

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

SciTech Connect (OSTI)

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

Curtis Smith; Steven Prescott; Tony Koonce

2014-04-01T23:59:59.000Z

3

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

Broader source: Energy.gov [DOE]

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

4

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

E-Print Network [OSTI]

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

5

Applications of Living Fire PRA models to Fire Protection Significance Determination Process in Taiwan  

SciTech Connect (OSTI)

The living fire probabilistic risk assessment (PRA) models for all three operating nuclear power plants (NPPs) in Taiwan had been established in December 2000. In that study, a scenario-based PRA approach was adopted to systematically evaluate the fire and smoke hazards and associated risks. Using these fire PRA models developed, a risk-informed application project had also been completed in December 2002 for the evaluation of cable-tray fire-barrier wrapping exemption. This paper presents a new application of the fire PRA models to fire protection issues using the fire protection significance determination process (FP SDP). The fire protection issues studied may involve the selection of appropriate compensatory measures during the period when an automatic fire detection or suppression system in a safety-related fire zone becomes inoperable. The compensatory measure can either be a 24-hour fire watch or an hourly fire patrol. The living fire PRA models were used to estimate the increase in risk associated with the fire protection issue in terms of changes in core damage frequency (CDF) and large early release frequency (LERF). In compliance with SDP at-power and the acceptance guidelines specified in RG 1.174, the fire protection issues in question can be grouped into four categories; red, yellow, white and green, in accordance with the guidelines developed for FD SDP. A 24-hour fire watch is suggested only required for the yellow condition, while an hourly fire patrol may be adopted for the white condition. More limiting requirement is suggested for the red condition, but no special consideration is needed for the green condition. For the calculation of risk measures, risk impacts from any additional fire scenarios that may have been introduced, as well as more severe initiating events and fire damages that may accompany the fire protection issue should be considered carefully. Examples are presented in this paper to illustrate the evaluation process. (authors)

De-Cheng, Chen; Chung-Kung, Lo; Tsu-Jen, Lin; Ching-Hui, Wu [Institute of Nuclear Energy Research, P. O. Box 3-3, Lung-Tan, Tao-Yuan Taiwan (China); Lin, James C. [ABSG Consulting Inc., 300 Commerce Drive, Suite 200, Irvine, CA 92602 (United States)

2004-07-01T23:59:59.000Z

6

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

SciTech Connect (OSTI)

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

7

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

SciTech Connect (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

8

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

E-Print Network [OSTI]

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

Norris, Ray

9

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

SciTech Connect (OSTI)

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

Robert Youngblood

2010-06-01T23:59:59.000Z

10

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

SciTech Connect (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

11

Multiple thermohaline states due to variable di usivity in a hierarchy of simple models.  

E-Print Network [OSTI]

of under-resolved FG models is constructed, the simplest of which is an eight-cell cube, to connect the two

Edwards, Neil

12

P&RA CoP Webinars  

Broader source: Energy.gov [DOE]

Topics of interest to the CoP have been identified with the help of the steering committee, and will be discussed in quarterly Webinars. 

13

Loss of spent fuel pool cooling PRA: Model and results  

SciTech Connect (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

14

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

SciTech Connect (OSTI)

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

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

1989-01-01T23:59:59.000Z

15

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

SciTech Connect (OSTI)

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

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

1993-09-07T23:59:59.000Z

16

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

Office of Environmental Management (EM)

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

17

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

Office of Environmental Management (EM)

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

18

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

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

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

19

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

Office of Environmental Management (EM)

ThomasKent Rosenberger Savannah River Remediation Stewart Walker EPA, HQ Cheryl WhalenDib Goswami State of Washington, Department of Ecology Ed Winner Commonwealth of Kentucky...

20

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

Office of Environmental Management (EM)

Modeling: Current Status and Future Applications (Alaa Aly, CHPRC and Dib Goswami, WA Ecology) 2:30 - 3:00 pm Discussion on Graded Approach (All) 3:00 - 3:15 pm Plans for Future...

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

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

Office of Environmental Management (EM)

National Laboratories Steve ThomasKent Rosenberger Savannah River Remediation Stewart Walker EPA, HQ Cheryl WhalenDib Goswami State of Washington, Department of Ecology Ed Winner...

22

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

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

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

23

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

E-Print Network [OSTI]

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

Kraan-Korteweg, Renée C.

24

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

Broader source: Energy.gov [DOE]

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

25

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

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

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

26

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

E-Print Network [OSTI]

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

Paris-Sud XI, Université de

27

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

E-Print Network [OSTI]

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

Siewert, Charles E.

28

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

E-Print Network [OSTI]

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

Karsai, Istvan

29

Estimation of effective diffusion coefficients in porous catalysts  

E-Print Network [OSTI]

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

Kulkarni, Shrikant Ulhas

2012-06-07T23:59:59.000Z

30

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

E-Print Network [OSTI]

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

Karsai, Istvan

31

Transport Analysis of Radial Electric Field in Helical Plasmas  

E-Print Network [OSTI]

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

32

HYDRODYNAMIC LIMITS FOR KINETIC EQUATIONS AND THE DIFFUSIVE APPROXIMATION OF RADIATIVE  

E-Print Network [OSTI]

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

Tzavaras, Athanasios E.

33

Energy, Climate & Infrastructure Security  

E-Print Network [OSTI]

-depthasneeded,withcontent ranging from Level 1 - 3 PRA and Fire PRA, to human reliability analysis, human factors for courses on nuclear reactor safety and Sandia-developed computational codes, used to model and simulate

34

Late Quaternary pollen records from Easter Island  

Science Journals Connector (OSTI)

... , K. The Mystery of Easter Island: The Story of an Expedition 165–199 (Sifton, Praed, London, 1919).

J. R. Flenley; Sarah M. King

1984-01-05T23:59:59.000Z

35

Books Received  

Science Journals Connector (OSTI)

... Islands Far Away. By Agnes G. King. Pp. xxvii + 256. (London: Sifton, Praed and Co., Ltd.) i8s.. net.

1920-11-04T23:59:59.000Z

36

Borderlands of Language in Europe:  

Science Journals Connector (OSTI)

... Historic Frontier of Christendom. By Vaughan Cornish. Pp. x + 105. (London: Sifton Praed and Co., Ltd., 1936.) 6s. net.

1937-12-11T23:59:59.000Z

37

Books Received  

Science Journals Connector (OSTI)

... . By Agnes G. King. Second edition. Pp. xxxii + 256. (London: Sifton, Praed and Co., Ltd.) 185. net.

1921-11-03T23:59:59.000Z

38

Training for Records and Information Management  

Broader source: Energy.gov [DOE]

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

39

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

E-Print Network [OSTI]

Nuclear Power Plants; NPPs). In 1975, the Atomic Energy Commission initiated the landmark Rasmussen study to BP to avert the oil spill in the Gulf of Mexico. In PRA research and applications, the nuclear [3] that led to the advent of PRA in the nuclear industry. Over the years, PRA has grown

Rubloff, Gary W.

40

Slide 1  

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

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

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

Evaluation of plastic–rubber asphalt: Engineering property and environmental concern  

Science Journals Connector (OSTI)

Abstract Waste rubber and plastic are accumulating heavily in China and causing significant environmental issues. In this study, the rubber and one kind of plastic, polypropylene (PP), were powdered and mixed with base asphalt to form plastic–rubber asphalt (PRA). This study evaluated PRA and the mixture in two-folds, engineering properties and environmental concerns. SBS asphalt, one commonly used asphalt binder in China, was adopted as a control. To evaluate the environmental burdens of PRA mixture and SBS asphalt mixture, a cradle-to-gate life cycle assessment (LCA) modeling was performed. Throughout the study, it is revealed that: (1) as a binder, PRA is weaker in terms of softening point, elastic recovery and fatigue performance versus SBS asphalt, but the differences are limited; (2) PRA mixture are close to SBS asphalt mixture for the high temperature, low temperature performances and water durability; (3) PRA mixture is more environmental-friendly compared to SBS asphalt mixture.

Bin Yu; Liya Jiao; Fujian Ni; Jun Yang

2014-01-01T23:59:59.000Z

42

The Poetic Impression of Natural Scenery  

Science Journals Connector (OSTI)

... Scenery. By Dr.VaughanCornish. Pp. vii + 90 + 4 plates. (London: Sifton Praed and Co., Ltd., 1931.) 6s. net. L. C. W ...

L. C. W. B.

1931-10-17T23:59:59.000Z

43

Books Received  

Science Journals Connector (OSTI)

... . Mrs. .Sro.res.bv Routledge. Pn. xxi4- Ar-4. (London:. Sifton, Praed, and Co., Ltd.)- yi's.fid. net. ...

1919-11-20T23:59:59.000Z

44

London on the Thames: a Study of the Natural Conditions that Influenced the Birth and Growth of a Great City  

Science Journals Connector (OSTI)

... Political Science: Geographical Studies, No. 3.) Pp. xiv + 189. (London: Sifton, Praed and Co., Ltd., 1923.) 7s. 6d. net.

1924-05-31T23:59:59.000Z

45

National Parks: and the Heritage of Scenery  

Science Journals Connector (OSTI)

... the Heritage of Scenery. By Dr.VaughanCornish. Pp. xi + 139. (London: Sifton Praed and Co., Ltd., 1930.) 5s. net.J.R.J. ...

J. R.

1930-09-13T23:59:59.000Z

46

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

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

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

47

Foreign Users | Stanford Synchrotron Radiation Lightsource  

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

at Customs) Passport Legal Permanent Residents (LPR) (Also known as Permanent Resident Aliens, PRA's, andor "green card" holders) Or Conditional Permanent Residents (CPR)...

48

Information Collection Management | Department of Energy  

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

49

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

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

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

50

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

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

Commission Washington, DC 20555-0001 SUBJECT: COMPARISON OF INTEGRATED SAFETY ANALYSIS (ISA) AND PROBABILISTIC RISK ASSESSMENT (PRA) FOR FUEL CYCLE FACILITIES Dear Chairman...

51

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

Broader source: Energy.gov [DOE]

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

52

J. Eckert, E.-M. Biermann-Ratjen, D. Höger (Hrsg.) Gesprächspsychotherapie. Lehrbuch für die Praxis  

Science Journals Connector (OSTI)

Zu einer Zeit, in der der Bundesausschuß für Ärzte und Krankenkassen über die sozialrechtliche Zulassung der Gesprächspsychotherapie als Richtlinienverfahren entscheidet, erscheint ein neues „Lehrbuch für die Pra...

H. Kächele

2007-03-01T23:59:59.000Z

53

September 2008 Memo on CPMS | Department of Energy  

Office of Environmental Management (EM)

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

54

CPMS Tables | Department of Energy  

Office of Environmental Management (EM)

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

55

List of International Projects for FY 2012 | Department of Energy  

Office of Environmental Management (EM)

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

56

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

SciTech Connect (OSTI)

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

Weiss, A.J. (comp.)

1987-02-01T23:59:59.000Z

57

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

SciTech Connect (OSTI)

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

Weiss, A.J. (comp.)

1988-02-01T23:59:59.000Z

58

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

SciTech Connect (OSTI)

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

59

Idaho National Laboratory Advanced Test Reactor Probabilistic Risk  

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

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

60

Augmenting Probabilistic Risk Assesment with Malevolent Initiators  

SciTech Connect (OSTI)

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

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

June 3, 2014 Webinar- Features, Events, and Processes: Practical Considerations for Development and Selection of Scenarios  

Broader source: Energy.gov [DOE]

P&RA CoP Webinar - June 3, 2014 - Features, Events, and Processes: Practical Considerations for Development and Selection of Scenarios Geoff Freeze (SNL) and Roger Seitz (SRNL)

62

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

E-Print Network [OSTI]

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

Moulin, Maria Teresa

2010-01-01T23:59:59.000Z

63

Probabilistic Risk Assessment for dairy waste management systems  

E-Print Network [OSTI]

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

Leigh, Edward Marshall

2012-06-07T23:59:59.000Z

64

NUREG-0668 MASTER* TITLE LIST PUBLICLY AVAILABLE DOCUMENTS THREE...  

Office of Scientific and Technical Information (OSTI)

Atoale Ena-gg Coaa 7903190733 Forward* Aaand 12 to Hear containing roipsnsa* to AEC qua*tic DCYOUNO. R. C. Atoaic Energy Coaatiai h 1979) 7J11J Atoaic Energ* CoaaLIon (Pra...

65

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

66

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

Office of Environmental Management (EM)

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

67

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

68

Applications of the EBR-II Probabilistic Risk Assessment  

SciTech Connect (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.

1993-01-01T23:59:59.000Z

69

Applications of the EBR-II Probabilistic Risk Assessment  

SciTech Connect (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

70

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

Broader source: Energy.gov [DOE]

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

71

MINISTRIO DA EDUCAO UNIVERSIDADE FEDERAL DO PARAN  

E-Print Network [OSTI]

-graduação em Direito Praça Santos Andrade, 50 ­ 3º Andar Tel.:(41)3310-2685 e 3310-2739 www SELE��O Comissão de Seleção: Prof. Dr. Eroulths Cortiano Junior e Prof. Dr. Rodrigo Xavier Leonardo CI�NCIAS JURÍDICAS Programa de Pós-graduação em Direito Praça Santos Andrade, 50 ­ 3º Andar Tel.:(41

Paraná, Universidade Federal do

72

A mathematical model of the productivity index of a well  

E-Print Network [OSTI]

. Hence, @ uw@t = @ u @t . The divergence theorem implies that @ u @t = 1 V Z Ludx = 1V Z w @u @~ dS: (4.2) Consequently, @ uw@t = qV , from which (4.1) easily follows. Let u1(x) be the solution to the auxiliary steady-state problem: Lu1 = 1V (4..., let u(x;t) = qu1(x) qV t. By virtue of the divergence theorem, Z w @u @~ dS = q: Consequently, u is a solution of the initial boundary value problem I with the initial distribution f1(x) = qu1(x). The di usive capacity J(u;t) on u(x;t) is constant...

Khalmanova, Dinara Khabilovna

2004-09-30T23:59:59.000Z

73

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

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

74

SAPHIRE 8 Volume 3 - Users' Guide  

SciTech Connect (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). 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

75

Peer Review of NRC Standardized Plant Analysis Risk Models  

SciTech Connect (OSTI)

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

76

The Zion integrated safety analysis for NUREG-1150  

SciTech Connect (OSTI)

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

77

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

SciTech Connect (OSTI)

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

Vierow, Karen; Aldemir, Tunc

2009-09-10T23:59:59.000Z

78

A comparison of experimental results and theoretical predictions for the rotordynamic coefficients of short (L/D = 1/6) labyrinth seals  

E-Print Network [OSTI]

CLEARANCE (mm) SEAL CLEARANCE (mm) Figure 11. Inlet tangential velocity vs. seal clearance for teeth-on-stator and teeth-on-rotor seals. 28 Swirl = 1 Pra 0. 45 tc =12000 cpm 0. 20 + 0. 18 A ~ 0. 1S E 014 D. 12 O 0. 10 M 0. 08 0. 08 1 = seal gl... CLEARANCE (mm) SEAL CLEARANCE (mm) Figure 11. Inlet tangential velocity vs. seal clearance for teeth-on-stator and teeth-on-rotor seals. 28 Swirl = 1 Pra 0. 45 tc =12000 cpm 0. 20 + 0. 18 A ~ 0. 1S E 014 D. 12 O 0. 10 M 0. 08 0. 08 1 = seal gl...

Pelletti, Joseph Michael

2012-06-07T23:59:59.000Z

79

Nuclear weapon system risk assessment  

SciTech Connect (OSTI)

Probabilistic risk assessment (PRA) is a process for evaluating hazardous operations by considering what can go wrong, the likelihood of these undesired events, and the resultant consequences. Techniques used in PRA originated in the 1960s. Although there were early exploratory applications to nuclear weapons and other technologies, the first major application of these techniques was in the Reactor Safety Study, WASH-1400, {sup 1} in which the risks of nuclear power accidents were thoroughly investigated for the first time. Recently, these techniques have begun to be adapted to nuclear weapon system applications. This report discusses this application to nuclear weapon systems.

Carlson, D.D.

1993-11-01T23:59:59.000Z

80

Curriculum Laboratory Drama: Instructional Planning  

E-Print Network [OSTI]

Handbook for the Actor 1986 T 792.028 Pra Gr. 10-12 Fundamentals of Voice and Diction (8th ed.) S 808.5 May for the Theater 3rd ed.: A Handbook of Teaching an Directing Techniques T 792.028 Spo Gr. 10-12 A Practical

Seldin, Jonathan P.

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

The School of Social Sciences Degree Offered: BA  

E-Print Network [OSTI]

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

Richards-Kortum, Rebecca

82

UNIVERSIDADE FEDERAL DO PARAN Setor de Cincias Jurdicas  

E-Print Network [OSTI]

-graduação em Direito - Mestrado e Doutorado Praça Santos Andrade, 50 - 3º andar - CEP 80.020-300 Curitiba Santos Salles Graça Egon Bockmann Moreira Direito Constitucional 7,7 Classificado 9. Bruno Cortez Torres Rodrigo Xavier Leonardo Direito Civil 7,5 Classificado 11. Cesar Felipe Bolzani Cesar Antonio Serbena

Paraná, Universidade Federal do

83

Praa Santos Andrade, n 50, Trreo Tel: 41-3310-2677 e-mail: npj@ufpr.br  

E-Print Network [OSTI]

Praça Santos Andrade, nº 50, Térreo Tel: 41- 3310-2677 ­ e-mail: npj@ufpr.br Centro, Curitiba SANTOS 9 GRR20081733 EDUARDA DE SOUSA LEMOS 10 GRR20081935 FABIANA MASSAKO NAKATANI 11 GRR20082221. RODRIGUES 19 GRR20104445 RAFAEL BORGES PINTO 20 GRR20084135 RENAN GUEDES SOBREIRA 21 GRR20053580 RODRIGO

Paraná, Universidade Federal do

84

Praa Santos Andrade, n 50, Trreo Tel: 41-3310-2677 e-mail: npj@ufpr.br  

E-Print Network [OSTI]

Praça Santos Andrade, nº 50, Térreo Tel: 41- 3310-2677 ­ e-mail: npj@ufpr.br Centro, Curitiba SANTOS 9 GRR20081733 EDUARDA DE SOUSA LEMOS 10 GRR20081935 FABIANA MASSAKO NAKATANI 11 GRR2008222120084135 RENAN GUEDES SOBREIRA 20 GRR20053580 RODRIGO LEAL COELHO 21 GRR20084607 THYAGO VARGAS FERREIRA 22

Paraná, Universidade Federal do

85

Commissioner George Apostolakis U.S. Nuclear Regulatory Commission  

E-Print Network [OSTI]

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

Bernstein, Joseph B.

86

HAI\\DS-ONMETEOROLOGY Stories,Theories,and Simple Experiments  

E-Print Network [OSTI]

were drafts ftom the ave of drc Creek god Aeolu. In Sicyon, an altar wd raised to praGe winds, md many particles (atoms) in a small space. When the space was large and the nlmbo of partides was snall

Short, Daniel

87

SAPHIRE 8 Volume 7 - Data Loading  

SciTech Connect (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

88

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

E-Print Network [OSTI]

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

Jo, Moon-Ho

89

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

SciTech Connect (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

2006-07-01T23:59:59.000Z

90

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

SciTech Connect (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

91

Towards a 21st Century Postal Service John C. Panzar  

E-Print Network [OSTI]

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

Bustamante, Fabián E.

92

Assessing Power Substation Network Security and Survivability: A Work in Progress Report1  

E-Print Network [OSTI]

infrastructure systems whose disruption would have an enormous impact on all of our lives. One of the most critical infrastructure systems identified was the electric power grid since this system supports all other], Probability Risk Assessment (PRA) [7], a prototype expert system [3], and a set of checklists from power

Krings, Axel W.

93

Risk Analysis and Probabilistic Survivability Assessment (RAPSA): An Assessment Approach for Power Substation Hardening1  

E-Print Network [OSTI]

or infrastructure system, and cause widespread fear from a major or prolonged service disruption [8]. In assessingRisk Analysis and Probabilistic Survivability Assessment (RAPSA): An Assessment Approach for Power System Analysis (SSA) with Probability Risk Assessment (PRA). The method adds quantitative information

Krings, Axel W.

94

Comparison of Integrated Safety Analysis (ISA) and Probabilistic Risk  

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

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

95

Information Collection Management | Department of Energy  

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

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

96

PAPERWORK REDUCTION ACT OF 1995  

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

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

97

Public Outreach  

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

98

Slide 1  

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

99

Probabilistic risk assessment of N Reactor using NUREG-1150 methods  

SciTech Connect (OSTI)

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

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

1989-11-01T23:59:59.000Z

100

Expert opinion in risk analysis: The NUREG-1150 methodology  

SciTech Connect (OSTI)

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

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

1988-01-01T23:59:59.000Z

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

Cable Hot Shorts and Circuit Analysis in Fire Risk Assessment  

SciTech Connect (OSTI)

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

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

1999-05-19T23:59:59.000Z

102

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

SciTech Connect (OSTI)

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

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

1984-01-01T23:59:59.000Z

103

Information Management and Supporting Documentation | Department of Energy  

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

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

104

Step by Step Instructions  

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

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

105

Site Screening, Site Selection,  

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

106

Generalized qudit Choi maps  

E-Print Network [OSTI]

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

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

2006-07-24T23:59:59.000Z

107

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

SciTech Connect (OSTI)

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

108

EDITAL n 007/2013 PPGD RESULTADO DA SEGUNDA ETAPA DA SELEO DOUTORADO 2013  

E-Print Network [OSTI]

e Doutorado Praça Santos Andrade, 50 - 3º andar - CEP 80.020-300 Curitiba ­ Paraná ­ Brasil Fone FAVORÁVEL 23 Rodrigo Eduardo Camargo Direito Civil José Antônio Peres Gediel FAVORÁVEL 24 Rodrigo Valgas dos Santos D. Administrativo Romeu Felipe Bacellar FAVORÁVEL 25 Samia Moda Cirino D. Trabalho Aldacy Rachid

Paraná, Universidade Federal do

109

UNIVERSIDADE FEDERAL DO PARAN Setor de Cincias Jurdicas  

E-Print Network [OSTI]

-graduação em Direito - Mestrado e Doutorado Praça Santos Andrade, 50 - 3º andar - CEP 80.020-300 Curitiba Classificado 7. Aulus Luiz Santos Salles Graça Egon Bockmann Moreira Favorável Classificado 8. Bruno Cortez Rodrigo Xavier Leonardo Favorável Classificado 10. Cesar Felipe Bolzani Cesar Antonio Serbena Favorável

Paraná, Universidade Federal do

110

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

SciTech Connect (OSTI)

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

111

Uncertainty associated with probabilistic prediction of nutrient transport by runoff  

E-Print Network [OSTI]

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

Jain, Mohit

1996-01-01T23:59:59.000Z

112

Level III probabilistic risk assessment for N Reactor  

SciTech Connect (OSTI)

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

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

1990-04-01T23:59:59.000Z

113

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

SciTech Connect (OSTI)

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

114

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

SciTech Connect (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

115

Risk management at the Oak Ridge National Laboratory research reactors  

SciTech Connect (OSTI)

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

116

Advanced neutron source reactor probabilistic flow blockage assessment  

SciTech Connect (OSTI)

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

Ramsey, C.T.

1995-08-01T23:59:59.000Z

117

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

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

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

1991-07-01T23:59:59.000Z

118

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

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

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

1993-08-01T23:59:59.000Z

119

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

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

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

1994-05-01T23:59:59.000Z

120

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

SciTech Connect (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

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

Developing and evaluating distributions for probabilistic human exposure assessments  

SciTech Connect (OSTI)

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

Maddalena, Randy L.; McKone, Thomas E.

2002-08-01T23:59:59.000Z

122

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

SciTech Connect (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 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

123

Risk-Informed Safety Margin Characterization Methods Development Work  

SciTech Connect (OSTI)

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

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

2014-09-01T23:59:59.000Z

124

Calcium sensitivity determinations by neutron activation analysis as applied to bone  

E-Print Network [OSTI]

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

Blasdel, Michael John

2012-06-07T23:59:59.000Z

125

The GTE Ceramic Recuperator for High Temperature Waste Heat Recovery  

E-Print Network [OSTI]

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

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

1984-01-01T23:59:59.000Z

126

A generalized land use study of the San Jacinto River watershed of Texas  

E-Print Network [OSTI]

Figure 1. Basic land Resource Areas LEGEND ; ? L B la ck la nd Pra i r ies CO Cocs t Prai r ie FC Forested Coastal Plain FC -C Forested Coastal Plain (Flatwoods) BO Bo t t om lands OT.'TSIDE HEAVY SOLID LI2IE - Boundary of the San Jacinto... Pi ? ? ft o ] 00 I to jco jco j ? co 03 ? 5 ^ O aS ?? ?? ?p U Pi ? ? ft O 4? CQ ? U O aS ?? ?* 43 U Pi ? ? ft O BG uo CM n CM c> o 2 0 t - cr...

Buckley, Frank A.

1951-01-01T23:59:59.000Z

127

Adaptive Sampling using Support Vector Machines  

SciTech Connect (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

128

Salinity Control in Irrigation Agriculture.  

E-Print Network [OSTI]

management pra TEST - DON'T GUESS. : .: Apply water uniformly by using a properly designed irrigation system and by leveling where necc Apply enough water for the crop plus enough to keep salt leached to a satisfactory level. A preplant irrigation may... be desirable. Irrigate more often than necessary under non-saline conditions. Provide adequate drainage. The free water table should be at least 5 to 6 feet below the SI Select crops tolerant to your salt conditions. Plant good seed under optimum moisture...

Longenecker, Donald E.; Lyerly, Paul J.

1957-01-01T23:59:59.000Z

129

Risk-based methods applicable to ranking conceptual designs  

SciTech Connect (OSTI)

In Ginichi Taguchi`s latest book on quality engineering, an emphasis is placed on robust design processes in which quality engineering techniques are brought ``upstream,`` that is, they are utilized as early as possible, preferably in the conceptual design stage. This approach was used in a study of possible future safety system designs for weapons. As an experiment, a method was developed for using probabilistic risk analysis (PRA) techniques to rank conceptual designs for performance against a safety metric for ultimate incorporation into a Pugh matrix evaluation. This represents a high-level UW application of PRA methods to weapons. As with most conceptual designs, details of the implementation were not yet developed; many of the components had never been built, let alone tested. Therefore, our application of risk assessment methods was forced to be at such a high level that the entire evaluation could be performed on a spreadsheet. Nonetheless, the method produced numerical estimates of safety in a manner that was consistent, reproducible, and scrutable. The results enabled us to rank designs to identify areas where returns on research efforts would be the greatest. The numerical estimates were calibrated against what is achievable by current weapon safety systems. The use of expert judgement is inescapable, but these judgements are explicit and the method is easily implemented on an spreadsheet computer program.

Breeding, R.J.; Ortiz, K. [Sandia National Labs., Albuquerque, NM (United States); Ringland, J.T. [Sandia National Labs., Livermore, CA (United States); Lim, J.J. [Lim and Orzechowski Associates, Alamo, CA (United States)

1993-11-01T23:59:59.000Z

130

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

SciTech Connect (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

131

Preliminary Hazards Analysis Plasma Hearth Process  

SciTech Connect (OSTI)

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

132

Risk Management for Sodium Fast Reactors.  

SciTech Connect (OSTI)

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

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

2015-01-01T23:59:59.000Z

133

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

SciTech Connect (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 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

134

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

SciTech Connect (OSTI)

Probabilistic Risk Assessment (PRA) and Human Reliability Assessment (HRA) are important technical contributors to the United States (U.S.) Nuclear Regulatory Commission’s (NRC) risk-informed and performance based approach to regulating U.S. commercial nuclear activities. Furthermore, all currently operating commercial NPPs in the U.S. are required by federal regulation to be staffed with crews of operators. Yet, aspects of team performance are underspecified in most HRA methods that are widely used in the nuclear industry. There are a variety of "emergent" team cognition and teamwork errors (e.g., communication errors) that are 1) distinct from individual human errors, and 2) important to understand from a PRA perspective. The lack of robust models or quantification of team performance is an issue that affects the accuracy and validity of HRA methods and models, leading to significant uncertainty in estimating HEPs. This paper describes research that has the objective to model and quantify team dynamics and teamwork within NPP control room crews for risk informed applications, thereby improving the technical basis of HRA, which improves the risk-informed approach the NRC uses to regulate the U.S. commercial nuclear industry.

Jeffrey C. JOe; Ronald L. Boring

2014-06-01T23:59:59.000Z

135

Probabilistic risk assessment for salt repository conceptual design of subsurface facilities: A techical basis for Q-list determination  

SciTech Connect (OSTI)

Subpart G ''Quality Assurance'' of 10 CFR Part 60 requires that the US Department of Energy (DOE) apply a quality assurance program to ''all systems, structures, and components important to safety'' and to ''design and characterization of barriers important to waste isolation.'' In April 1986, DOE's Office of Geologic Repositories (OGR) issued general guidance for formulating a list of such systems, structures, and components---the Q-list. This guidance called for the use of probabilistic risk assessment (PRA) techniques to identify Q-list items. In this report, PRA techniques are applied to the underground facilities and systems described in the conceptual design report for the Salt Repository Project (SRP) in Deaf Smith County, Texas. Based on probability and dose consequence calculations, no specific items were identified for the Q-list. However, evaluation of the analyses indicated that two functions are important in precluding off-site releases of radioactivity: disposal container integrity; and isolation of the underground facility by the heating, ventilation, and air conditioning (HVAC) systems. Items related to these functions are recommended for further evaluation as the repository design progresses. 13 refs., 20 figs.

Chen, C.P.; Mayberry, J.J.; Shepherd, J.; Koza, H.; Rahmani, H.; Sinsky, J.

1987-12-01T23:59:59.000Z

136

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

SciTech Connect (OSTI)

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

137

SAPHIRE 8 Software Project Plan  

SciTech Connect (OSTI)

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

138

EPRI/NRC-RES fire human reliability analysis guidelines.  

SciTech Connect (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

139

Notices  

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

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

140

Notices  

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

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

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141

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)

142

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)

143

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

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

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

144

Physics-Based Stress Corrosion Cracking Component Reliability Model cast in  

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

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

145

DOE F 4200.33.cdr  

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

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

146

Page not found | Department of Energy  

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

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

147

Page not found | Department of Energy  

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

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

148

Physics-Based Stress Corrosion Cracking Component Reliability Model cast in  

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

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

149

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

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

150

Page not found | Department of Energy  

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

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

151

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

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

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

152

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

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

153

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

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

154

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

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

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.

155

Revised accident progression and risk analyses for NUREG-1150  

SciTech Connect (OSTI)

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

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

1987-01-01T23:59:59.000Z

156

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

SciTech Connect (OSTI)

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

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

1989-09-01T23:59:59.000Z

157

NUREG-1150 risk assessment methodology  

SciTech Connect (OSTI)

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

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

1987-01-01T23:59:59.000Z

158

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

SciTech Connect (OSTI)

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

159

Performing Probabilistic Risk Assessment Through RAVEN  

SciTech Connect (OSTI)

The Reactor Analysis and Virtual control ENviroment (RAVEN) code is a software tool that acts as the control logic driver and post-processing engine for the newly developed Thermal-Hydraulic code RELAP-7. RAVEN is now a multi-purpose Probabilistic Risk Assessment (PRA) software framework that allows dispatching different functionalities: Derive and actuate the control logic required to simulate the plant control system and operator actions (guided procedures), allowing on-line monitoring/controlling in the Phase Space Perform both Monte-Carlo sampling of random distributed events and Dynamic Event Tree based analysis Facilitate the input/output handling through a Graphical User Interface (GUI) and a post-processing data mining module

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

2013-06-01T23:59:59.000Z

160

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

SciTech Connect (OSTI)

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

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

2013-05-01T23:59:59.000Z

Note: This page contains sample records for the topic "usive pra ctices" from the National Library of EnergyBeta (NLEBeta).
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161

Entangling Power of Permutations  

E-Print Network [OSTI]

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

Lieven Clarisse; Sibasish Ghosh; Simone Severini; Anthony Sudbery

2005-02-07T23:59:59.000Z

162

Risk Informed Safety Margin Characterization Case Study: Selection of  

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

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

163

Wyoming's Economic Future: Planning for Sustained Prosperity  

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

164

ORISE: Radiation Emergency Medicine - Continuing Medical Education Course  

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

165

Integration of Advanced Probabilistic Analysis Techniques with Multi-Physics Models  

SciTech Connect (OSTI)

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

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

2014-07-30T23:59:59.000Z

166

Optimized Uncertainty Quantification Algorithm Within a Dynamic Event Tree Framework  

SciTech Connect (OSTI)

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

J. W. Nielsen; Akira Tokuhiro; Robert Hiromoto

2014-06-01T23:59:59.000Z

167

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

SciTech Connect (OSTI)

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

168

Conversion of Questionnaire Data  

SciTech Connect (OSTI)

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

2011-01-01T23:59:59.000Z

169

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.

170

Natural convection heat transfer analysis of ATR fuel elements  

SciTech Connect (OSTI)

Natural convection air cooling of the Advanced Test Reactor (ATR) fuel assemblies is analyzed to determine the level of decay heat that can be removed without exceeding the melting temperature of the fuel. The study was conducted to assist in the level 2 PRA analysis of a hypothetical ATR water canal draining accident. The heat transfer process is characterized by a very low Rayleigh number (Ra {approx} 10{sup {minus}5}) and a high temperature ratio. Since neither data nor analytical models were available for Ra < 0.1, an analytical approach is presented based upon the integral boundary layer equations. All assumptions and simplifications are presented and assessed and two models are developed from similar foundations. In one model, the well-known Boussinesq approximations are employed, the results from which are used to assess the modeling philosophy through comparison to existing data and published analytical results. In the other model, the Boussinesq approximations are not used, thus making the model more general and applicable to the ATR analysis.

Langerman, M.A.

1992-05-01T23:59:59.000Z

171

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

SciTech Connect (OSTI)

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

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

1989-01-01T23:59:59.000Z

172

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

SciTech Connect (OSTI)

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

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

1993-12-01T23:59:59.000Z

173

Developing new theoretical models of the formation of atomic collision cascades and subcascades in irradiated solids  

SciTech Connect (OSTI)

A new theoretical model is developed for the investigation of atomic collision cascades and subcascades in irradiated solids consisting of atoms of a single type. The model is based on an analytical description of the elastic collisions between moving atoms knocked out of the crystal lattice sites and the immobile atoms of the lattice. The description is based on the linear kinetic Boltzmann equation describing the retardation of primary recoil atoms (PRAs) in irradiated solids. The laws of conservation for the total number and the kinetic energy of moving atoms, which follow from the kinetic Boltzmann equation, are analyzed using the proposed model. An analytical solution is obtained for the stationary kinetic Boltzmann equation, which describes the retardation of PRAs for a given source responsible for their production. A kinetic equation for the moving atoms and the corresponding laws of conservation are also analyzed with allowance for the binding energy of atoms at the crystal lattice sites. A criterion for determining the threshold energy of subcascade formation in irradiated solids is formulated. Based on this criterion, the threshold energy of subcascade formation is calculated using the Thomas-Fermi potential. Formulas are presented for determining the mean size and number of subcascades formed in a solid as functions of the PRA energy.

Metelkin, E. V.; Ryazanov, A. I., E-mail: ryazanoff@comail.ru; Semenov, E. V. [Russian Research Center Kurchatov Institute (Russian Federation)

2008-09-15T23:59:59.000Z

174

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

E-Print Network [OSTI]

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

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

2015-01-01T23:59:59.000Z

175

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

SciTech Connect (OSTI)

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

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

2014-08-01T23:59:59.000Z

176

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

SciTech Connect (OSTI)

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

177

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

SciTech Connect (OSTI)

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

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

1990-01-01T23:59:59.000Z

178

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

SciTech Connect (OSTI)

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

Freels, J.D.

1993-01-01T23:59:59.000Z

179

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

SciTech Connect (OSTI)

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

Freels, J.D.

1993-07-01T23:59:59.000Z

180

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

SciTech Connect (OSTI)

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

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

1997-02-01T23:59:59.000Z

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181

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

SciTech Connect (OSTI)

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

Plys, Martin; Burelbach, James; Lee, Sung Jin; Apthorpe, Robert [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)] [Fauske and Associates, LLC, 16W070 83rd St., Burr Ridge, IL, 60527 (United States)

2013-07-01T23:59:59.000Z

182

Development of Simplified Probabilistic Risk Assessment Model for Seismic Initiating Event  

SciTech Connect (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

183

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

SciTech Connect (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

184

Experimental EPR-Steering of Bell-local States  

E-Print Network [OSTI]

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

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

2009-09-04T23:59:59.000Z

185

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

186

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

SciTech Connect (OSTI)

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

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

2012-11-01T23:59:59.000Z

187

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

SciTech Connect (OSTI)

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

188

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

SciTech Connect (OSTI)

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

189

Effects of improved modeling on best estimate BWR severe accident analysis  

SciTech Connect (OSTI)

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

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

1984-01-01T23:59:59.000Z

190

On the safety of aircraft systems: A case study  

SciTech Connect (OSTI)

An airplane is a highly engineered system incorporating control- and feedback-loops which often, and realistically, are non-linear because the equations describing such feedback contain products of state variables, trigonometric or square-root functions, or other types of non-linear terms. The feedback provided by the pilot (crew) of the airplane also is typically non-linear because it has the same mathematical characteristics. An airplane is designed with systems to prevent and mitigate undesired events. If an undesired triggering event occurs, an accident may process in different ways depending on the effectiveness of such systems. In addition, the progression of some accidents requires that the operating crew take corrective action(s), which may modify the configuration of some systems. The safety assessment of an aircraft system typically is carried out using ARP (Aerospace Recommended Practice) 4761 (SAE, 1995) methods, such as Fault Tree Analysis (FTA) and Failure Mode and Effects Analysis (FMEA). Such methods may be called static because they model an aircraft system on its nominal configuration during a mission time, but they do not incorporate the action(s) taken by the operating crew, nor the dynamic behavior (non-linearities) of the system (airplane) as a function of time. Probabilistic Safety Assessment (PSA), also known as Probabilistic Risk Assessment (PRA), has been applied to highly engineered systems, such as aircraft and nuclear power plants. PSA encompasses a wide variety of methods, including event tree analysis (ETA), FTA, and common-cause analysis, among others. PSA should not be confused with ARP 4761`s proposed PSSA (Preliminary System Safety Assessment); as its name implies, PSSA is a preliminary assessment at the system level consisting of FTA and FMEA.

Martinez-Guridi, G.; Hall, R.E.; Fullwood, R.R.

1997-05-14T23:59:59.000Z

191

Human Reliability Analysis for Small Modular Reactors  

SciTech Connect (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

192

Evaluation of potential severe accidents during low power and shutdown operations at Grand Gulf, Unit 1: Evaluation of severe accident risks for plant operational state 5 during a refueling outage. Supporting MELCOR calculations, Volume 6, Part 2  

SciTech Connect (OSTI)

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