ML20126E069

Regulatory and Technical Reports (Abstract Index Journal). Compilation for Third Quarter 1992,July-September

ML20126E069
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Issue date:	11/30/1992
From:	NRC OFFICE OF ADMINISTRATION (ADM)
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References
NUREG-0304, NUREG-0304-V17-N03, NUREG-304, NUREG-304-V17-N3, NUDOCS 9212280293
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NUREG-0304 Vol. :17, No. 3 Regulatory and Technical Reports (Abstract Index Journal)

Compilation for Third Quarter 1992 July - September U.S. Nuclear Regulatory Commission Omce of Adtr: ?stration

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1 NUREG--0304 Vol.17, No. 3 Regulatory and Technical Reaorts (Abstract Index Journal)

Compilation for Third Quarter 1992 July -- September Date Put.lished: November 1992 Regulatory Publications liranch ~

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CONTENTS Preface. . ... . ..... .. .. . .. ......... . .... .. . .. ... . . . . . . . . . . . -v Index' Tab ,

Main Citations and Abstracts .,, ... . . .... . . . . . . . . . . , . . . . . . '1 e Staff Reports

• Conference Proceedings

• Contractor Reports

• International Agreement Reports Secondary Report Number index . . . . . . . . . . . . . , . . . . . . . . . . . . . . . . . . . . 2--

Personal Authorindex ...... .. . .. ,, . . , , . . . . . . . . . . . . . . . , . . :3 S u bject i nde x . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . , , . , , , 4 NRC Originating Organization Index (Staff Reports) . . . . . . . . . . . . , , , . . . . . . . 5

- NRC Originating Organization index (International Agreements) . , , , , . . . . . . . . . . . . . . . . . 6 NRC Contract Sponsor index (Contractor Reports) . ' . , . . . . . . . . . . . . . . . . .. ., -7

. Contractor Index . . . . . ........ . . . .. .. . . . . . , , . . . . . . , . . . . . . . . . . . . 8 International Organization index . . . . . .. . , , , . . . . .. . . . , , . .. . . . . . Licensed Facility index . . . . . . . . . ..... . . . . . . . . . , , . , . . . . . . . . . . . . . . . . 10-n,,

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PREFACE This compilation consists of bibliograpNc data and abstracts for the formal regulatory and technical reports issued by the U.S. Nuclear Regulatory Commission (NRC) Staff and its contractors. It is NRC's intention to publish this compilation quarterly and to cumulate it annually. Your comments will be ap-preciated. P! ease send them to:

Technical Publications Section Regulatory Publications Branch Division of Freedom of Information and Publications Services P 223 U.S Nuclear Regulatory Commission Washington, D.C. 20555 _

The main citations and abstracts in this compilation are listed in NUREG number order: NUREG-XXXX, NUREGfCP XXXX, NUREG/CR XXXX, and NUREG/lA XXXX. These precede the following indexes:

Seconday Report Number Index Personal Author Index Subject index NRC Originating Organization Index (Staff Reports)

NRC Originating Organization index (h temational Agreements)

NRC Contract Sponsor index (Contractor Reports)

Contractor index international Organization index Licensed Facility Index A detailed explanation of the entries precedes each iridex The bibliographic elements of the main citations are the following:

Staff Report NUREG-0808i MARK 11 CONTAINMENT PROGRAM EVALUATION AND ACCEPTANCE CRITERIA. "

ANDERSON, C.J. Division of Safety Technology. August 1981. 90 pp. 8109140048. 09570:200.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organizational unit of '

author, (5) date report was published, (6) number of pages in the report, (7) the NRC Ds:ument Control <3 '

L-System accession number, (8) the microfiche address (for intemal NRC use).

- Conference Report NUREG/CP-0017: EXECUTIVE SEMINAR ON THE FUTURE ROLE OF RISK ASSESSMENT.AND

- RELIABILITY ENGINEERING IN NUCLEAR REGULATION. JANERP, J.S. Argonne National

- Laboratory. - May 1981,141 pp. 8105280299. ANL-81-3. 08632:070.

Where the entries are (1) report number, (2) report title, (3) report author, (4) organization that compiled -

the proceedings, (5) date report was published, (6) number of pat;es in the report, (7) the NRC Docu-ment Control System accession number, (8) the report number of the originating organization, (9) the microfiche address (for NRC intem l use). _;

-Contractor Report NUREG/CR 1556: STUDY OF ALTERNATE DECAY HEAT REMOVAL CONCEPTS FOR LIGHT WATER REACTORS-CURRENT SYSTEMS AND PROPOSED OPTIONS. BERRY, D.L.; BENNETT, P.R.

Sandia Laboratories, May 1981.100 pp. 8107010449. SAND 80-0929. 08912:242.

Where the entries are (1) report nu'mber, (2) report title, (3) report authors, (4) organizational unit of :

- authors or publisher, (5) date report was published, (6) number _of pages in the report, (7) the NRC '

• Documer.t Control System accession number, (8) the report number of the originating organization (if -

given), and (9) the microfiche address (for NRC intemal use),

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Iriternational Agreement Report NUREG/lA-0001: ASSESSMENT OF TRAC PD2 USING SUPER CANNON AND HDR EXPERIMENTAL DATA. NEUMANN, U. Kraftwerk Union. August 1986. 223 pp. 8608270424, 37659:138.

Where the entries are (1) report number, (2) report title (3) report author, (4) organizational unit of author, (5) date report was published, (6) number of pages in the report, (7) the NRC Document Control System accession number, (8) the report number of the originating organization (if given), and 19) the microfiche address (for NRC intemal use).

The following abbreviations are used to identify the document status of a report:

ADD - addendum APP - appendix ~

DRFT - draft ERR - errata N - number R - revision S - supplement V - volume Availability of NRC Publications Copies of NRC staff and contractor reports may be purchased either from the Government Printing Office (GPO) or from the National Technical information Service, Springfield, Virginia 22161. To purchase documents from the GPO, send a check or money order, payable to the Superintendent of Documents, to the following address:

Supenntendent of Documents U.S. Govemment Printing Office Post Office Box 37082 Washington, DC 20013-7082 You may charge any purchase to your GPO Deposit Account, MasterCard charge card, or VISA charge card by calling the GPO on (202)275-2060 or (202)275-2171. Non-U.S. customers must make payment h advance either by International Postal Money Order, payable to the Superintendent of Documents, c:

by draft on a United States or Canadian bank, payable to the Superintendent of Documents.

NRC Report Codes The NUREG designation, NUREG-XXXX, indicates that the document is a formal NRC staff generated report Contractor-prepared formal NRC reports carry the report code NUREG/CR-XXXX. This type of identification replaces contractor-established codes such as ORNL/NUREG/TM-XXX and TREE-NUREG-XXXX, as well as various other numbers that could not be correlated with NRC sponsorship of the work being reported.

In addition to the NUREG and NUREG/CR codes NUREG/CP is used for NRC-sponsored conference proceedings and NUREG/1A is used for international agreement reports.

All these report codes are controlled and assigned by the staff of the Publishing and Translations Section of the NRC Divisian of Publications Services.

vi

- Moln. Citations and Abstracts The report listings in this compilation are arranged by report number, where NUREG XXXX is -

an NRC staff originated report, NUREG/CP-XXXX is an NRC-sponsored conference report, NUREG/CR XXXX is an NRC contractor-prepared report, and NUREG/lA XXXX is an inter-national agreement report The bibliographic information (see Preface for details) is followed by a brief abstract of this report.

1 NUHEG-0040 V16 NO2: UCENSEE CONTRACTOR AND NUREG-0386 D06 p?* UNITED STA1ES NUCLEAR REGULA-VENDOR INSPECTION STATUS REPORT. Quarterly TORY COMMIS' A .TAFF PRACTICE AND PROCEDURE -l Report.Apnl - June 1992 (White Book)

• Divison of Reactor in- O! GEST.Commiss. ppeal Board And Licensing Board -- 1 J

specton & Safeguards (870411921003). July 1992.129pp, Decisions. July 1972 oepternber 1991.

• Office of the General  !

9209020331.62939 252 Counsel (Post 860701). August - 1992. 596pp. 9209170066.

I This penodcal covers the results of inspections performed by 63143.001.

the NRC's Vendor inspecbon Branch that have been distnbuted This 3rd revision of the sixth editeon of the NRC Practice and to the inspected organaabons dunny the penod from Aprit Procedure Digest contains a dgest of a number of Commission, through June 1992. Atomec Safety and Ucensing Appeal Board, and Atomic Sr/ety and Licensing Board decisions issued dunng the penod of July NUREG-0090 V15 N01: REPORT TO CONGRESS ON ADNOR, 1,1972 to September 30,1991, interpreting the NRC's Rules of MAL OCCURRENCES, January March 1992.

• Office for Ana9 sas & Evaluahon of Operat onal Data. Duector. July 1992.30pp. Pracuca in 10 CFR Part 2.

9206060263. 62647:335. NUREG-0525 V01: SAFEGilARDS.

SUMMARY

EVENT LIST Secton 208 of the Energy Reorgancaton Act of 1974 identi- (SSEL). Pre-NRC Through December 31, iO39. YARDUMIAN.J.;

hen an abnormal occurrence as an unscheduled incident or FADDEN.M. Division of Safeguards . & Transportation (Post event that the Nuclear Regulatory Commiss on determines to be 870413). July 1992. 47Dpp. 9209170124. 63145:001.

signilicant from the standpoint of pubhc health and safety and The Safeguards Summary Event Ust provides brief summa-requees a quarterly report of such events to be made to Con, ties of hundreds of safeguards-related events involving nuclear gross This report covers the period January through March matenal or facilities regulated by the U.S. Nucisar Regulatory 1992. Three abnorma! occurrences involving medical therapy Commiss on. Events are described under the categones: bonib-misadministratens at NRC-hcensed facihties are discussed in related, intrusion, missing / allegedly sto;en, transportaborvrelat-this report. There were no abnormal occurrences at a nuclear ed, tampering / vandalism, arson, fsrearms related, radeological power plant, and none were reported by NRC's Agreement States. The report also contains informabon updahng some pro, sabotage, non-radiological sabotage, and miscellaneous. De-cause of the public 6nterest, the mascellaneous section also in-viously reported abnormai occurrences.

ciudes events reported involving source matonal, byproduct ma-NUREG-0090 V15 NO2: REPORT TO CONGRESS ON ABNOR. terial, and natural uranium, which are exempt from safeguards MAL OCCURRENCES.Apnt-June 1992.

• Office for Analysis & requirements. Information in the event desenptions was ob.

Evaluanon of Operational Data, Director. September 1992. - tainod from officia! NRC reports.

27pp. 9210130169. 63462:036.

Section 208 of the Energy Recrgancation Act of 1974 idenu- NUREG-0525 V02: SAFEGUARDS

SUMMARY

EVENT UST fies an abnormal occurrence as an unscheduled incident or (SSEL). January 1, 1990 Through December 31, 1991, event that the Nuciear Regulatory Commission determines to be YAROUM1AN Ja 8'ADDEN.M. Division of Safeguards & Trans-sgnificant from the standpoent of public health and safety and portation - (Post 8704t3) : July 1992. 174pp.- 9209170120, requires a quarterty report of such events to be made to Con- 63147:001.

gress. This report covers the penod April through June 1992, See NUREG-05:5,V01 abstract. '

Three abnormal occurrences invoMng medical therapy misad-minstrations and one invoMng a medical diagnostic misadmir> NUREG 0540 V14 N05: TITLE UST OF DOCUMENTS MADE istration at NRC-hcensed facilities are discussed in this report. PUBUCLY AVAILABLE.May 131,1992.* Division of Freedom There was one abnormal occurrence at a nuclear power plant- of information & Publicatons Services (Post 890205). July 1992.

and none were reported by NRC's Agreement States. The 565pp. 9206060272. 62651:001.

report also contains information updating some previously re- This docun ent is a monthly pubhcation containing descrip-ported abnormal occurrences. tions of information received and generated by the U.S. Nuclear -

Regulatory Commission (NRC). Thts information includes: (f) .

l NUREG 0304 V17 Not: REGULATORY AND TECHNICAL RE- docketed matenal associated with civihan nuclear power plants PORTS (ABSTRACT INDEX JOURNAL) Compilation For First Quarter 1992, January-March

• Dmsion of Freedom of Informa- and other uses of radioactive materials; and (2) nondocketed tion & Pubhcations Services (post 890205). Jure 1992. 41pp. material received and generated by NRC pertinent to its role as a regulatory agency. The following indexes are included: Per-9207270307. 62506.242. sonal Author, Corporate Source, Report Number, and Cross This Journal includes all formal reports in the NUREG senes Reference of Enclosures to Principal Documents.

prepared by the NRC staff and contractors; proceedings or ron-ferences and workshops; as well as international agreiement re-ports. The entries in this compilahon are indexed for access by NUREG-0540 V14 N06: TITLE UST OF DOCUMENTS MADE PUBUCLY AVAILABLE. June 1-30, 1992.

• Division of Freedom btle and abstract, secondary report number, personal author, of informatson & Pub 6 cations Services (Post 890205). August

. subject, NRC organcation for staff and intematonal agree-ments, conbactor, intemational organizabon, and licensed facih- 1992. 304pp 0209020320. 62938:001; See NUREG-0540,Vf 4,N05 abstract.

ty.

1

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2 -Main Citations and Abstracts NUREG-0540 V14 N07: TITLE LIST OF DOCUMENTS MADE This report covers the major activities, events, decisions and PUBLIC: Y AVAILABLEJuly 1 31, 1992.

• Divison of Freedom planrung that took place during fiscal year 1991 within the U S, of Information & Pubhcations Services (Post 890205). Septem. Nuclear Regulatory Commission (NRC) or involving the NRC.

ber 1992. 349pp. 9209220359. 63242:001, See NUREG-0540,V14,N05 abstract. NUREG 1214 R10: HISTORICAL DATA

SUMMARY

OF THE SYS-TEMATIC ASSESSMENT OF LICENSEE PERFORMANCE.

NUREG4750 V35 N05: NUCLEAR HEGULATORY COMMISSION ALLENSPACH,F. Dmson of Licensee Performance & Quality ISSUANCES FOR MAY 1992.Pages 189-203.

• Dvision of Evaluaten (870411-921003). August 1994 127pp.9209170112.

Fraedom of Information & Publicabons Services (Post 890205). 63140.076.

July 1992. 21pp. 9208060257. 62654:221-The Historical Data Summary of the Systematic Assessment -

Legalissuances of the Commission, the Atom #c Safety and Li- of Licensee Performance (SALP) is produced periodically by the consing Board Panel, the Administrative Law Judges, and NRC U.S. Nuclear Regulatory Commission. This summary provides Program Offices are presented.

the results of ; a assessment for each facility by NRC regiott it -

NUREG 0750 V35 N06: NUCLEAR REGULATORY COMMISSION is further divided into the following sections: Section 1 presents ISSUANCES FOR JUNE 1992.Page 205-260.

• Division of Free. the rnost recent SALP report ratings for facilities in operation dom of information & Pubhcatons Services (Post B90205). and under construction; Secton 2 presents a chronological list-August 1992 63pp. 9209020326. 62939:189. ing of all SALP report ratings for each operahng facility, Secten See NUREG-0750,V35,N05 abstract. 3 presents a chronological listing of all SALP report ratings for each facility under constructon. For histoncal purposes, past NUREG-0837 V12 N02: NRC TLD DIRECT RADIATION MONI- construction ratngs for facihbes that recently have been h-TORING NETWOr1K. Progress Report Apnl4une 1992.

censed also are hsted in Section 3.

STRUCKMEYER,R.; MCNAMARA.N. Region 1 (Post 820201).

September 1992. 232pp. 9209240250. 63279:046. NUREG 1242 V01: NRC REVIEW OF ELECTRIC POWER RE-This report provides the status and results of the NRC Ther. SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR molun.inescent Dosimeter (TLD) Drect Radiawn Monitonng UTILITY REQUIREMENTS DOCUMENT. Program Summary.

• Network, ft presents the radiation levels measured in the vicincy Associate Director for Advanced Reactors & License Renewal of NRC hconsed facihties throughout the country for the second (Post 910918). August 1992.138pp. 9209240187. 63278.268.

quarter of 1992. The staff of the U.S. Nuclear Regulatory Comrnission has pre-NUREG-0933 S14: A PRIORITIZATION OF GENERIC SAFETY "" "l 8 ** " "

• Y

'SSUES. EMRIT,R.; RIGGS.R.; MILSTEAD.W.; et at Division of w f ac M esearch Inse s had' @

Safety issue Resolution (Post 880717). Au9ust 1992. 82pp. *# en s Ment - Wam h mary,o to document the results of its review of the Electric r esents he prionty rankings for genenc safety U

ea Ins a paw @ WaW ham W

-issues related to nuclear power plants. The purpose of these #

rankings is to assist in the timely and efficierst allocation of NRC f the overall purpose and scope of the Requirements Docu-resources for the resolution of those safety issues that have a ment, the background of the staff's review, the review approach significant potential for reducing risk. The safety pnonty rankings W m su aM a em W m W W WW are HIGH, MEDIUM, LOW, and DROP and have been assignea **

on the basis of nsk sgnificance estimates, the ratio of risk to NUREG-1242 V02 P01: NRC REVIEW OF ELECTRIC POWER costs and other impacts estimated to result if resolutions of the RESEARCH INSTITUTE'S ADVANCED LIGHT WATER REAC.

saiety issues were implemented, and the consideration of un- TOR UTILITY REQUIREMENTS DOCUMENT. Evolutonary certainties and other quanttatve or qualitabvo factors. To tne Plant Designs.Chaptw 1,

• Associste Director for Advanced Re-extent pracbcal, estimates are quanttative.

actors & License Renewal (Post 910918). August 1992.513pp.

NUREG-0936 VII N02: NRC REGULATORY AGENDA.Ouarterly 9209240193. 63277:084.

Report,Apnl4une 1992.

• Division of Freedom of information & The staff of the U.S. Nuclear Regulatory Commission has pre-Publications Services (Post 690205: July 1992. 151pp. pared Volume 2 (Parts 1 and 2) of a safety evaluation report-9208250253, 62884:001, (SER), "NRC Review of Electric Power Research Institute's Ad-The NRC Regulatory Agenda is a compilation of all rules on venced Ught Water Reactor Utility Requirements Document ,

which the NRC has recently completed action, of has proposed Evolutionary Plant Designs," to document the rer its of its action, or is considering action, and all petitjons for rulemaking revtew of the Electne Power Research Insttute's " Advanced .

which have been received by the Commsssion and are pending Light Water Reactor Uulity Requirements Document." This SER .

disposttion by the Commissiort The Regulatory Agenda is up- gives the results of the staff's review cf Volume 11 of the Re-dated and issued each quarter, quirements Document for evolutionary ptant designs, which con-sists of 13 chapters and contains utility design requirements for NUREG-0940 Vit N02: ENFORCEMENT ACTIONS: SIGNIFl* an evolutiont.ry nuclear power plant (approximately 1300 CANT ACTIONS RESOLVED.Quarterty Progress Report.Apnl- megawatts - electric).

June 1992.

• Ofc of Enforcement (Post 870413). August 1992.

451pp 9209170097. 63141:001. NUREG-1242 V02 P02: NRC REVIEW OF ELECTRIC POWER This compdabon summarizes significant enforcement actions RESEARCH INSTITUTE'S ADVANCED LIGHT WATER REAC-that have been resolved during one quarterfy period (April . TOR UTILITY REQUIREMENTS DOCUMENT. Evolutionary June 1992) and includes copies of letters, Notices, and Orders Plant Designs. Chapters 213.

• Associate Director for Advanced sent by the Nuclear Regulatory Commission to licensees with Reactors & License Renewa! (Post 910918). August 1992.

respect to these enforcement &ctions. It is anbcapated that the 610pp.9209240198.63274:001.

information in this publication will be widely disseminated to See NUREG-1242,V02,P01 abstract.

managers and employees engaged in activities licensed by the NRC, so that actions can be taken to improve safety by avoid- NUREG 1272 V06 N01: OFFICE FOR ANALYSIS AND EVALUA-ing future violations similar to those described in this publica. TION OF OPERATIONAL DATA.1991 Annual Report - Power -

' Reactors.

• Office for Ana!ysis & Evaluation of Operational Data, Director. July 1992. 244pp. 9209020312. 62938:305.

l NUREG 1145 V08: U.S. NUCLEAR REGULATORY COMMISSION The annual report of the U.S. Nuclear Regulatory Commis-

l. 1991 ANNUAL REPORT.

• Otfice of Administration (Post sion's Office for Anatysis and Evaluation of Operational Data i 890205). July 1992. 270pp. 9208260241, 62882:001. (AEOO) is devoted to the activities performed during 1991. The i

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Main Citations and Abstracts 3 report is pubfished in two separate parts NUREG-1272, Vol. 6. Executive Director for Operabons. the Director, Ofice of Nu:le--

No.1, covers power reactor.s and presents an overview of the ar Regulatory Research, or to the Derectos Office of Nudear operst#g exponence of the nudear power industry from the Mateful Safety and Safeguards All reports prepared by the NRC purspective, inc*udng comments about (te trends of some Con.mittee have been made available to the pubhc through the key performance measures. The report also includes the pnnc6- NRC Pubhc Documeat Room and the U S. Library of Congress.

pai findings and essues identhed in AEOD studios over the past ysar and summarizes information from such sources as Imensee NUREG 1442 R01: EMERGENCY RESPONSE ' RESOURCES GUIDE.For Nuclear Power Plant Emergencies. WEINSTEIN.E.;

event reports, dmgnostic evatustions, and reports to the NRC's BATES,G. NRC No Uetakd Affiliation Given.

• Federal Emer-Operabons Center. The reports contain a discumon of the incb dont invostigaton Team program and summarize the incident gency Management Agency. Jufy 1992. 43pp. 9208240333.

FEMAAEP 17. 62666:077, Investt9ation Team and Augmented inspechon Team reports for On August 28 and September 18.1990, the States of Louisk that group of icensees Ni) REG 1272, Vol 6, No. 2. covers ar.a and Mississippi, Gutt States Utihties, five local panshes, six nonreactors and presents a review of the events and concems Federal agencios, and the Amercan Nuclear insurers participat-dunng 1991 associated with the use of heensed ma',onal in non.

reactor appleations, such as porsonnel overexposures and ed in a post emergency TABLE-TOP exercise in Baton Rouge, Louimana. One of the products developed from that exponence medical misadministrabons Each votume contains i irst of the AEOO reports issued for 1984 1990. is this guide for understanding the responsibilities and obtaining NUREG-1272 V06 N02: OFFICE FOR ANALYSIS AND EVALUA, or za th r1 o n TION OF OnERATONAL DATA 1991 Annual Report - Nonreac* ment. This first revision of that guide broadens the focus of the tors.

• Othee for Anatysis & Evaluaton of Operatonal Data. D" onginal document. Also, new information defines the major Fed-rector. August 1992.143pp. 9209170074. 63142:092, eral response facilities. This guide should assist State and local The annual report of the U.S. Nuclear Regulatory Commis* government organizations with identfying and obtaining those ston's Office for Analyses and Evaluation of Operational Data resources for the post emergency response .when their re-(AEOD) is devoted to the activit es performed during 199L The sources have been exhausted.

report es published in two separate parts. NUREG 1272, Vol 6 No 1, covers power reactors cnd presents an overview of the NUREG-1451: STAFF TECHNICAL POSITION ON INVESTIGA-operating exponence of the nuclear power industry from the TIONS TO IDENTIFY FAULT DISPLACEMENT HA2ARDS AND NRC perspective, including comments about the trends of some SEISMIC HAZARDS AT A GEOLOGIC REPOSITORY.

key performance measures. The report also includes the princi- MCCONNELL K.L; BLACKFORD,M.E.: IBRAHIM,A.B. Divison of pal f ridings and issues identf*>d in AEOD studies over the past High-Level Waste Management (Post 870413).. July 1992,68pp.

year and summarizes informat on from such sources as licensee 9209170131. 63140:203.

event reports, diagnost'c evaluations, and reports to the NRC's 10 CFR Part 60 does not specify the manner in which poten-Operabons Center. The reports contain a discusseon of the inci- tial fault displacement hazards and seismic hazards at a candi-dont investigaton Team program and summarize the incident date site for a geolog.c repository are to be idenhfied. The pur-Investiganon Team and Augmented Inspection Team reports for pose of this staff technical position (STP), therefore, is to pro-that group of licensees. NUREG-1272, Vol 6. No. 2, covers vide guidance to the U S. Department of Energy (DOE) on ac-nonreactors and presents a review of the events and concerns ceptable geologic repository investigations that can be used to dunng 1991 associated with the use of hcensed matenalin non- identfy fault displacement hazards and seismic hazards. The reactor appheahons, such as personnel overexposures and staff considers that the approach this STP takes to investiga.

medical misadministrabons. Each volume contains a list of the bons of fault displacement and seismic phenomer.a is approprv AEOD reports issued for 1981-1991. ate for the collection of sufficient data for input to anafraes of fault displacement hazards and seismic hazards, both for the NUREG-1377 R03: NRC RESEARCH PROGRAM ON PLANT preciosure and postclosure pe.formance periods. However, de-AGING. LISTING AND SUMMARIES OF REPORTS ISSUED taded analyses of fault displacement and seismic data, such as THROUGH JULY 1992. KONDIC,N.N. Dmaron of Engineenng those required for detailed assessments of repository perforrn-(Post 870413). September 1902. 108pp. 9210130165- ance, may identify the need for additional investigations. Section 63462:065. 2.0 of this STP describes the 10 CFR Part 60 requirements that The U.S Nuclear Regulatory Commission is conductng the form the basis for inveshgations to describe the fault displace-Nuclear Plant Aging Research (NPAR) Program. This is a com- ment hazards and seismic hazards at a geologic repository.

prehensive hardware- onented engineenng research program Technical pc* tion statements and corresponding discussons

• focused on understanding the agmg mechanisms of compo. are presented in Sections 3.0 and 4.0, respectivety. Technical nents and systems in nuclear power plants. The NPAR program positon topics in this STP are categorrzed thusly- (1) investiga-also focuses on methods for simulating and monitonng the tion considerations; (2) Investigations for fault displacement haz-aging- related degraonhon of these components and systems. ards, and (3) investigation for seismic hazards. .

in addition, it provides recommendatons for effective maints.

nance to manage aging and for the impicmentation of the re- NUREG-1457: RESOURCES AVAILABLE FOR NUCLEAR search results in the regulatory process. This document cor" POWER PLANT EMERGENCIES UNDER THE PRICE ANDER-tains a hsting and index of reports generated in the NPAR pro- SON ACT AND THE ROBERT T. STAFFORD DISASTER gram that were issued through July 1992 and summaries of REUEF AND EMERGENCY ASSISTANCE ACT.WEINSTEIN,E.

those reports. Each summary describes the elements of the re- Division of Operational Assessment (Post 870413). July 1992.

search covered in the report and outlines the significant results 23pp. 9208170156. 62791:295.

I For the convenience of the user, the reports are indexed by Through a series of TABLETOP exercises and other events perse.,nal author, corporate author, and subject. that involved participation by State and Federal organizations, the need was identihed for further explanation of financial and NUREG-1423 V03: A COMPILATION OF REPORTS OF THE AD

! VISORY COMMITTEE ON NUCLEAR WASTE. July 1991 June other related resources available to individuals and State and 1992.

• Advisory Committee on Nuclear Waste. August 1992. local govemments in a major emergency at a nuclear power

. 8 f pp 9209f 70f 05. 63140-002. plant. A group with representabves from the Nuclear Regulatory This compitabon contains 19 reports issued by the Advisory Commission, the Federal Emergency Management Agency, and l Committee ott Nuciear Waste (ACNW) dunng the fourth year of the American Nuclear insurers / Mutual Atomic Energy Liabihty sts opernhon. The reports were submitted to the Chairman and Underwnters was estabhshed to work toward tfus end. This l report is the result of that effort.

Commissoners of the U S. Nuclear Regulatory Commission, the

--._-_,..-.__,._._...,__.._,__.,m ..-m..,,,.-..._-, . .m,. .,___,.,,,._.._,_m.. .

p. ~~ _ . - . - - - - - - - - - - - - _ - . . - - - - . - . - - , - - .

4- Main Citations and Abstracts i

NUREG-1465 DRFT FC:-.. ACCIDENT SOURCE TERMS FOR power plant aging from programs sponsored by the Offica of )

LIGHT WATER NUCLEAR POWER PLANTS. Draft Report For Nuclear Regulatory Research, U.S. Nuclear Regulatory Comma-Comment. SOFFER,L.; BURSON.S B.; FERRELLO.M ; et al. D- von. The conference also provided an opportanity for engnieem vision of Safety issue Resolution (Post 890117) Jne 1992. ed scientists from around the wond to exchange technical ir-46pp 9208100123. 62680:298 formahon and discuss future lnternatonal ccoperation. The in 1962 the U.S. Atomic Energy Commircion pubashed TiO- papers and talks appear in the order in which they were pro-14344,"Calculat+on of Distance Factors for Power and Test Re. sented at the conference, and they are grouped by technical j actors" which spectfied a release of hssion products from the sestron. ,

core to the reactor containmont in the event nt a postulated ac- I cident involving " substantial mettdown of the core". This NUREQ/CP-0122 V02; PROCEEDINGS OF THE AGlNG RE-

" source term", the basis for the NRC's %gulatory Gudes 1.3 SEARCH INFORMATION CONFERENCE, BERANEK,A.F. Dw 1 and 1.4, has been used to determine comohance with the sion of Engineering (Post 870413), September 1992. 465,)p.

NRUs reactor site cnteria,10 CPR Part 100, and to evaluate 9209290380. 63336:031.

other important plant performance toquirements During the See NUREGICP-0122,V01 abstract.

- past 30 years substantial additional info ro on fission prod. NUREG rCP-0123: PROCEEDINGS OF THE SECOND NRC/

uct releases has been developed based on significant severe ASME SYMPOSIUM ON PUMP AND VALVE TESTING. Held At accident research. This document utrzes this research by pro- The Hyatt Regency HotelWashmgton.DC, July 21 23. 1992,

• viding more reakstic estimates of the " source term" release mto EG&G idaho. Inc, July 1992. 562pp. 02u7270274 EGG-2676.

containment, in terms of timing, nuchde types, quantities, and 62510:001, chemical form, given a severe core melt accident. Th,s revised The 1992 Symposium on Pump and Valve Testing, }ointly

' source term' is to be applied to the design of future hght water sponsored by tho Board on Nuclear Codes and Standards of reactors (LWRs) Current LWR licensees may voluntarily pro- the American Socief of Machanical Engineers and by the No-pose soplications based upon it. These will be reviewed by the cleu Regulatory Comrnission, provides t forum for the discus-NRC staff.

sion of current o:ograms and methods for inservice testing and NtJ sEG 1470 V0-1: CHIEF FINANCIAL OFFICER'S ANNUAL motor operated valve testing at nuclear power plants. Tft3 sym.

REPORT - 1992.

• Office of the Controllst (Post 690205) Sep, posium also provides an oppnrtunity to discuss the need to inw tomber 1992 28pp. 9210060022. 63424 307. prove that testing in order ts netp ensure the reliable perform-The Chief Financial Officers Act of 1990 requires the NRC ance of pumps and vatas. The participation of industry repre.

Chief Financial Officer to prepare and submit an annual report sentativos, regulators, and consultants results in the discussion to the agency head and the Director of the Office of Mansge- of a broad spectrum of ideas and perspectives regarding the im-ment and Budget. This 1992 report is the first annual report for provement of inservice testing of pumps and valves at nuclear the NRC and includes a description and analysis of the status oower plants.

of financial management and a summary of the reports or, inter-nal accounting and administrative control systems' NUREG/CP-0124: WORKSHOP ON THE USE OF PRA METHOD-OLOGY FOR THE ANALYSIS OF REACTOR EVENTS AND NUREG/CP-0120: PROCEEDINGS OF THE FIFTH WORLHOP OPERATIONAL DATA. RASMUSON.DAt Dtvision of Safety

ON CONTAINMENT INTEGRITY, Held in Washington.DC.May Programs (Post 870413). OtNGMAN,S, Sandia National Labora-12-14,1992, P ARKS,M B. HUGHEY.C.E. Sandia National Lab- tones. June 1992,133pp. 9208070150. 62676'055.

oratoriet Jufy 1992. 654pp. 9208170148. SAND 92-0173. A workshop entitled The Use of PRA Methodology for the 62790:001. Arialysis of Reactor Events and Operational Data" was held on The Fifth Workshop on Containment Integnty was held in January 29 30, 1992 in Annapolis, Maryland. Over 50 partich Washington, DC, on May 12 14, 1992. The purpo*e of these pants from the NRC, its contractors. and others participated in workshops ss to provide an intemational forum for trL exchange the meetings Dunng the first day, presentations were made by of information on performance of containments in nuclear power invited speakers to discuss issues in twvant topics. On the

plants under severe accident loadings. Severe accident investi- second day, discussion groups were held to focus on three gahons of eusting containment designs as well as future ad- areas

(1) risk significance of operational events; (2) industry

_ vanced containments were presented dunng the workshop. nsk profile and genenc concems, and (3) nsk monitoring and There were 145 participants at the workshop from 15 countnes risk based performance indicators. Summaries of the discussion Ivan Sehn, Chairman of the NRC, provided the opening address sessions are contained in the report as well as important in-for the meeting. A total of 39 papers were presented on the foi- sights gained from the discussions.

towing topics: Containment Design Considerations for Severe Accident Conditions, Advanced Containment Designs and Relat. NUREG/CR 2907 VIO: RADIOACTIVE MATERIALS RELEASED ed Research. Contatnment Behavior Under Accident Conditions, FROM NUCLEAR POWER PLANTS. Annual Report 1989.

Testing / Analysis of Containment Systems, and Containment TICHLER J.; NORDEN.K4 CONGEMI,J. Brookhaven National Operational Experience (Leakage, Aging, and Operation). A Laboratory. September 1992. 300pp. 9210050068. BNL-

! copy of the final program, including last minute changes, is pro. NUREG 51581. 63384:238.

i vided in these proceedings. Papers that were presented at the Releases of radioactive matettais in airborne and hquid ef-wurkshop make up the body of this report The workshop was fluents from commercial hght water reactors during 189 have hosted by Sandia National Laboratories under the sponsorship been compiled and reported. Data on sohd waste shioments as of the U.S. Nuctear Regulatory Commission. Pnncipal organizers well as selected operating information have been included. This

.for the workshop were James F. Costello of the U S. Nuclear report supplements earlier annual reports issued by the former Regulatory Commission and Walter A. von Riesemann and M. Atomic Energy Commission and the Nuclear Regulatory Com-Brad Parks of Sandia National Laboratones. rnission. The 1989 release data are summarized in tabular form.

NUREG/CP-0122 - V01: PROCEEDINGS OF THE AGING RE-SEARCH INFORMATION CONFERENCE. BERANEK.A.F. Divi- NUREC/CR-3320 V02: LWR PRESSURE VESSEL ' SURVEIL +

ston of Engineenng (Post 870413) Septemoor 1992. 560pp. LANCE DOSIMETRY IMPROVEMENT PROGRAM. PSF Startup 9209290377. 63337;136. Expenments. MCELROY,W.N.; GOLD,Ra MCGARRY,E.D. Bat.

This report presents the proceedings of ;he Aging Research telle Memonal Institute. Paofic Northwest Laboratory. July 1992.

Information Conference held at the Holiday inn Crowne Plaza in 76pp, 9208260297. WHC-EP-0204. 62890.136.

Rockville, Maryland, on March 24-27, 1992. This conference The meta!!urgical irradiation experiment at the Oak .1dge Re-was held to disseminate ressarch results in the area of nuclear search Reactor Poolside Facihty (ORR-PSF) is one of the series d

l-s-___._______.____.___________.__

Main Citations and Abstracts 5 of benchmark expenments in the framework of the Light Water Transient Reactor Analysis Code (TRAC) for the U S. Nuclear Reactor Pressure Vessel Surveihance Dosimetry improvement Regulatory Commisson and the public. The TRAC-BF1/ MODI Program (LWR PV SDlP) The goal of this program is to test, version of the computer code provides a but estimate analysis against welbestabbshed benchmarks, the methodologms and capabbty for analyzing the full range of postulated accidents en data bases that are used to predict the irradiation embrittlement boihng water reactor (BWR) systems and related facihties. This and fructure toughness of pressure vessel and support structure version prowdes a consistent and untfied ana!ysis capabAty for steels The prediction methodology includes procedures for neu- analyzing all areas of a large- or small-break loss-of-coolant ac-trol physics calculatons. dos metry and spectrum adostment i cider't (LOCAL beginning wsth the blowdown phase and continu-methods metailurgical tests. and damage correlations. The ,ng through heatup, reficod with quenchtng- and, finalty, the refilt benchmark empenments serve to vakdate, improve, and stand-phase of the accident Also provided is a basic capabihty for the ardae these procedures lhe results of this program are imple~ anatysis of operatonal transients up to and including anticipated mented in a set of ASTM Standards on pressure vesset surved-lance procedures lhese, in turn, may be used as guidos for the transients without scram (ATWS). The TRAC-BF1/ MOD 1 ver.

g gg gg g g gg gg, nuclear industry and for the Nuclear Regulatory Comrnisuon (NRC). To serve as a bencnmark, a very careful charactenza- ment calculations using the two TRAC-BF1 versions show over-ton of the ORR-PSF expenment is necessary, both in terms of @ Ms m mew e e W mWe Ws neutron flux-fluence spectia and of metallurgical test results as compared to carber versions of .he TRAC-BWR series of Stat:stically determined uncer'arnties must be given in terms of compuwt codes.

vanances and covanances to malie compansons between pro-dicions and expenmental results meaningful Detailed descrip- NUREG/CR 4409 V04: DATA BASE ON DOSE REDUCTION RE-tons of the PSF phys 4cs-dosimetry startup expenments and SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

' KHAN,T A ; VULIN,D Sa LIANG.H.: et al. Brookhaven National

'" W Laboratory. August 1992 225pp. 9209220454. BNL-NUREG-NUREG/CR 3950 V0h FUEL PERFORMANCE ANNUAL 51934. 63238 001.

REPORT FOR 1989 BAILEY,W.Ja BERTING,F M. Battello Me- This is the fourth volume in a series of reports that provide rnonal Instaute, Pacibc Northwest Laboratory WU,S L Dmsion information on dose reduction research and health physics of Systems Technology (890827-921003) June 1992. 228pp- technology for nuclear power plants. The information is taken 9207270321. PNL.5210. frorn a data base maintained by Brookhaven National Laborato-62506 310'f desenption of fue! perform-This annual report provides a boe

,y s ALARA Center for the Nuclear Regulatory Commission.

anco dunng 1989 in commercial nuclear power plants. Onef This report presents information on 118 new or updated summanes of fuel design changes, fael suntemance programs-fuel operating expenence and trends, fuel problems, high- rotects, covering a wide range of acteties. Protects including burnup fuel exponence, and items of general significance are steam generator degradation, decontamination, rcbotics im-provided References to additional, more detailed informabon provements in reactor materials, and inspecton techniques, and related NRC evaluations are included. among others, are descnbed in the research section of the report. The section on health physics technology includes some NUREG/CR-4356 V01: TRAC-BF1/ MOD): AN ADVANCED s:mple and very cost effective projects to reduce radiation ex-BEST-ESTIMATE COMPUTER PROGRAM FOR BWR ACCI- posures Included in this volume is a detailed desenption of how DENT ANALYS:S Model Desenption. BORKOWSKl.J A.; to access the BNL data bases which store this informatKn. All WADE.N L; GILES,M M ; et at EG&G ldaho. Inc. August 1992- pro}act abstracts from this report, as well as many other ueful 306pp 9209220469. EGG-2626. 63224401, documents, can be accessed, with permission, tnrough our w The TRAC-BWR code development program at the Idaho Na- line system. ACE. A computer equipped mth a modem, or i. f ax tional Engineonng Laboratory has developed versions of the machine, is all that is required to connect to ACE. Many fea-Transient Reactor Anatyrs Code (TRAC) for the U.S. Nuclear tures of ACE, including so tware, hardware, and communica-Regulatory Commisuon and the puDlic. The TRAC-BFI/ MOD 1 ~

tions specifics, are explained in this report.

version of the computer code proudes a best- estimate analysis capaldty for analynng the full range of postulated accidents in NUREG/CR-4469 V13: NONDESTRUCTIVE EXAM l NATION boshng water reactor (BWR) systems and related f acihties. This (NDE) RELIABillTY FOR INSERECE INSPECTION OF LIGHT version provides a consistent and unified ana'ysis capability for WATER REACTORS Semiannual Heport. October 1990-March analyzing all areas of a large- or sma4 break loss-of coolant ac- 1991. DOCTOR,S R.; GOOD.M S.; HEASLER,P.G.; et al. Bat-cident (LOCA), beginning min the blowdown phase and continu-tetie Memonal Instnute, Pacific Northwest Laboratory July 1992.

ing through heatup, reflood eth quenching, and, finaHy, the refill BSpp. 9208250264. PNL-5711. 62870.001, phase of the accident. Also provided is a basic capab hty for the analysis of operational transients up to and including anticipated The Evaluation and improvement of NDE Reliabihty for in-service inspection of Light Water Reactors (NDE Rehabihty) transients without scram (ATWS). The TRAC-BF1/ MOD 1 ver-sion proouces results consistant with previous versions. Assess- Program at the Pacific Northwest Laboratory was estabhshed by ment calculations using the two TR.AC-BF1 versions show over- the Nuclear Regulatory Commission to determine the rehabhty allimprovements in agreement wth data and computat:on times of current inservice inspection OSI) techniques and to develop compared to earher versions of the TRAC-BWR senes of recomrnendations that mil ensure a suitably high inspection reli-abihty. The obl ectives of this program include determining the reliabihty of ISI performed on the pomary systems of commer.

NUREG/CR-4356 V02: TRAC-BF1/ MOO 1: AN ADVANCED BEST- cial hght+ater reactors (LWRs), using prooabilistic fracture me-ESTIMATE COMPUTEFi PROGRAM FOR BOILING WATER chanics ana!ysis to determme the impact of NDE unrehability on REACTOR ACCIDENT ANALYSIS User's Guide, RETTIG,W.H.; syuem safety; and evaluating rehabdity improvements that can WADE,N L. GILES,M.Ma et al EG&G Idaho, Inc. June 1992- be achieved wtth improved and advanced technology. A final 297pp 9208100136. EGG 2626. 62680 001.

obl ectrve is to formulate recommended revisions to ASME Code See NUREG/CR-4356.V01 abstract- aW Regulatory for,uirements, tased on material properties, NUREG/CR-4391: TRAC /BF1-MODI MODELS AND CORRELA. serwce conditions, and NDE uncertatnties. The program scope TIONS. BORKOWSKLJ A : WADE,N L; ROUHAN!,S Z..; et al. is kmited to ISt of the pnmary systems including the piping.

EG8G idaho, Inc. August 1992. 458pp. 9209220474 EGG. vessel, and other components inspected in accordance with 2680. 63239.001. Section XI of the ASME Code This is a progress report cover-The TRAC-BWR code development program at the Idaho Na- ing the programmatic work f om October 1990 through March tional Engineenng Laboratory has developed versions of the 1991-

. --. - . . .~ . _ ~ . . - - - . . . - ~ . . . _ . - . . - . ,

l l

V l

6' Main Citations and Abstracts NUREG/CR-4469 V14: NONDESTRUCTIVE EXAMINATION NUREG/CR-4674 V15: PRECURSORS TO POTENTIAL SEVERE

' (NDE) RELIABILITY FOR INSERVICE INSPECTION OF LIGHT CORE DAMAGE ACCIDENTS: 1991 A STATUS REPORT. Main

- WATER REACTORS. Semiannual Report.Apnl 1991. September Report And Appendix A. MiNARICK J.W. Science Applicabons 1991. DOCTOR,S.R.; DIAZ,A.A.; FRILEY,J R.; et at Battelle Me- Internabonal Corp. (formerly Science Applications, Inc.).

monal institute, Pacific Northwest Laboratory. July 1992. 82pp. CLETCHER.J.W.; COPINGER,0.A.; et at Oak Ridge National. i 9208250273. PNL-5711. 62871.228. - Laboratory. September 1992c 170pp. 9210070028. ORNL/

The Evaluaton and improvement of NDE Reliability for 1rw l NOAC-232. 63427:005. -

i service inspection of Light Water Reactors (NDE Reliability) Twenty-eight operatonal events with conditional probabihties Program at the Pacific Northwest Laboratory was established by of core damage of 1.0 X 10(.6) or higher occurring at commer-the Nuclear Regula'ory Commission to determine the rehability cial light- water reactors dunng 1991 are considered to be pre-of current inservica inspechon (ISI) techniques and to develop cursors to potential severe core damage. These are desenbed recommendabons that will ensure a suitably high inspection reli- along with associated significance estimates, categorization, ability. The objectives of this program include determining the and subsequent analyses. Ihts study is a conbnuation of earlier . j reliability of Isl performed on the pnmary systems of commer- work, which evaluated the 1969 to 1981 and 1984 to 1990 cial hght. water reactors (LWRs); using probabihsbc fracture me- events The report discusses (1) the general rationale for this charwcs anatysis to determine the impact of NDE unreliab#y on study, (2) the selechon and documentaten of events as precur.

system safety; and evaluabng reliability improvements that can sors. (3) the eshmaton and use of conditional probabihties of be achieved with improved and advanced technology. A final subsequent severe core damage to rank precursor events, and objectrve is to formulate recommended revisions to ASME Code (4) the plant models used in tha analysis process.

and Regulatory requirements, based on material properties, service conditions, and NDE uncertainties The program scope NUREG/CR-4674 V16: PRECURSORS TO POTENTIAL SEVERE is hmited to ISI of the pnmary systems including the p! ping. CORE t>AMAGE ACCIDE~NTS; 1991 A STATUS vessel, and other components inspected in accordance with REPORT. Appendices B,C, And D. MINARICK,J W. Science Ap-Section XI of the ASME Code. This is a progress report cover, plicatons international Corp. (formerty Science Applications, ing the programmatic work from Apni 1991 through Septerruer Inc.) CLETCHER J.W.; COPINGER D.A.; et al. Oak Ridge Na-1991, bonal Laboratory. Septembe 1992. 600pp. 9210070034.

ORNL/NOAC-232. 63425:003.

NUREG/CR-4599 V02 N1: SHORT CRACKS IN PIPING AND See NUREG/CR-4674,V15 abstract.

PIPING WELDS.Seniannual Report, April-September 1991.

WILKOWSKl,G.M.; BRUST,F.; FRANCINLR.; et al. Battelle Me- NUREG/CR 4744 V06 N1: LONG-TERM EMBRITTLEMENT OF monal Institute, Columbus Laboratones. September 1992. CAST DUPLEX STAINLESS STEELS IN LWR 21Spp. 9209280056. BMI-2173. 63335:001. SYSTEMS. Semiannual Report. October 1990 - March 1991.

This is the third semiannual report of the U.S, Nuclear Regu. CHOPRA,0 K. Argonne National Laboratory. August ' 1992.

latory CommissaWs Short Cracks in Piping and Piping Welds 199pp. 9209220414. ANL-91/22. 63241001, research program. This 4 year program began in March 1990. This progress report summarizes work performed by Argonne The overal! objective of this program is to venty and improve National Laboratory on long-term thermal embnttlement of cast fracture analyses for ctrcumferentally cracked large-diameter duplex stainless steels in LWR systems during the siir months nuclear piping with crack sizes typ6cally used in teak before- from October 1990 tc March 1991. Charpy-impact, tensile, and break analyses or inservice flaw evaluations. fracture toughness data are presented for several heats of cast stainless steel that were aged up to 58,000 h at ternperatures NUREG/CR 4667 V14: ENVIRONMENTALLY ASSISTED CRACK' of 294400* C. The results indicate that thermal agmg increases ING IN LIGHT WATER REACTORS, Semiannual Report.Oc,ouer the tensile stress and decreases the fracture toughness 1991 March '1992. CHUNG,H M.; of the materials. In general, CF-3 steels are the least sensitive KASSNER,T.F,; MAJUMDAR,5.; et al. Argonne National Labora- to thermal aging embnttlement and CF 8M steels are the most tory. August 1992. 65pp. 9209220422. ANL-92/30. 63240:263-sensitive. The increase in flow stress of fully aged cast stainless This report summanzes work performed by Argonne Natuna' Laboratory on fatigue and environmentally assisted cracking in steels is =10% for CF 3 steei> and = 20% for CF-8 and hght water reactors dunng the six months from October 1991 CF-8M steels. The fracture toughness J ic and average tear-ing modulus for heats that are sensitive to thermal aging through March 1992. Topics that have been investgated dunng (e g., CF 8M steels) are as low as = 90 kJ/ma and = 60,-

, this period include: (1) fatigue and stress corrosion cracking of respectively-low-alloy steel used in piping and in steam generator and reac-tor pressure vessels; (2) radiation-induced segregation and irra- NUREG/CR 4819 V02: AGING AND SERVICE WEAR OF SOLE.

diation-assisted SCC of Type 304 SS after accumulation of rela- NOtO-OPERATED VALVES USED IN SAFETY SYSTEMS OF hvely high fluence; and (3) update of a crack growth data base NUCLEAR POWER PLANTS. Evaluation Of Monitonng Methods.

for austenitic and femtic steels in high- temperature water. Ex- KRYTER,R.C. Oak Ridge National Laboratory, Jufy 1992 91pp, isting data on fabgue of low-alloy steel in LWR environments 9208260239. ORNL/TM-12038. 62878:006.

have been reviewed. Based on fracture-mechanics models and Solenoid. operated valves (SOVs) were studied at Oak Ridge engineering judgernent, interim fatigue deeign curves are oeing Nabonal Laboratory as part of the USNRC Nuclear Plant Aging

. developed that are consistent with available fatgue-hfe data. ML Research (NPAR) Program. The primary objechve of the study 4

crochemscal and rnicrostructural changes in high- and commer- was to identify, evaluate, and recommend methods for inspec-cial-purity Type 304 SS specimens from control-blade absorber tion, surveillance. monstonng, and maintenance of SOVs that tubes and a control-blade sheath from operating BWRs were can help ensure their operational readiness- that is, their ability studied by Auger electron spectroscopy and scanning electron to perform required safety functions under all anticipated oper-microscopy. Slow. strain-rate-tensHe tests were conducted on ir- atmg condibons, since fatture of one of these small and relative-radiated specimens in air and in simulated BWR water at 289 ty inexpensive devices could have serious consequences under degrees C. Cract growth data on fracture mechanics specimens certain circumstances. An earlier (Phase 1) NPAR program study of austenitic and femtec steels in simulated BWR water, devel- described SOV failure modes and causes and identified measur-oped in this program over the past 8 years, are compiled into a able parameters thought to be linked to the progression of ever-

- data base along with references that contain detatis of test present degradation mechanisms that may uturnately resutt in i methods, material compos 4 bons, metallographic information, and functional failure of the valve. Using this earlier work as a guide, comparisons of data with predicbons of Secton XI of the ASME the present (Phase 11) study focused on devising and Phen dem-Ch onstrating the effectveness of techniques and equipment with 1 .

o.v e.,-y-,, . . y- . 9 - wet-a o.- c ,,-7 -w ye mmg- r ,, w ,.y , =rwe=e- y w e,-yvr, rni - . , w ny w. evy mwe y w.p .y v v y +e e r es u see r w N' we.

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Main Citations and Abstracts -7

_l

' wNch to measure periormance parameters that show promise NUREG/CR 4832 V03 P1: ANALYSIS OF THE LASALLE UNIT 2 for detecting the presence and trending the progress of such NUCLEAR POWER PLANT: RISK METHODS INTEGRATION degradabons before they reach a cntical stage. Intrusive tech- AND EVALUATION PROGRAM (RMIEP)_Intemal Events Acci-niques requiring the add $on of magnetic or acoustic sensors or dent Sequence Ouantificaton. Main Report. PAYNEAC.; -

the apphcation of spectat test signals were investigated bnefly, DANtEL S W WHITEHEAD.D W.; et al. Sandia Natonal Labora-but maior emphasis was placed on the examinaten of condi- tones. Aug,ust 1992. 161pp. 9209220470. SAND 92-0537, tron-indicating techniques that can oe apphed with minimal cost -63244.020.

and impact on plant operauon. These include monitoring coil This volurne presents the methodology and results of the in-mean temperature remotely by means of coil dc resistance or ternal event accident sequence analysis of the LaSalle Unit 2 ac impedance, determirung valve plunger position by means of nuclear power plant performed as part of the Level 111 Probabi-cod ac impedance, venfying unrestricted SOV plunger move, list c Risk Assessment being performed by Sandia National Lab-ment by measuring current and voltage at their critical bistable oratories for the Nuclear Regulatory Commissert The total in-(puMn and drop-out) values, and detecting the presence of temal core damage frequency has a mean valve of 4.41E-05/R-shorted tums or insulaton breakdown within the solenoid coil yr with a 5th percentle of 2.05E-6/R-yr, a median value of using interrupted-current test methods. The first of these tech. 1.64E 05/R-yr, and a 95th percentde of 1.39E 04/R yr. The ruques, though perhaps the simplest conceptualty, will likely dominant sequences involve a loss of off site power (LOSP),im-benefit the nuclear industry most because SOVs have a history mediate or delayed failure of on-site AC power resulting in sta __

of failure in service as a result of unwitting operabon at exces. tron-blackout, and failure of the reactor core isolation coohng sive Umperatures. system (RCIC). The events most important to risk reduction are:

frequency of LOSP. non-recovery of offsite power within one NUREG/CR-4832 Vot: ANALYSIS OF THE LASALLE UNIT 2 NU- hour, diesel generator (DG) cooling water pump common mode CLEAR POWER PLANT: RISK METHOOS INTEGRATION AND failure, and non-recoverable isolation of RCIC <tunng station EVALUATION PROGRAM (RMIEP). Summary. PAYNE,A C. blackouts, The events most important to risk increase are: fail-Sandia Natonal Laboratones. July 1992.130pp. 9208060095. ure of various AC power circuit breakers resulting in partialloss SAND 92 0537. 62652:207. of onsite AC power, failure to scram, and DG cooling water This alume presents an overview of the methodology and re, comrnon mede failure The dominant contributors to uncertainty suits of the integrated accident soquence analysis (Level 1) of are: controt circuit failure rates, relay coil failure to energize, en-ergded relay coils faihng deenergized, frequency of LOSP, and the LaSalle Unit 2 nuclear power plant performed as part of the Level ill PRA pertormed by Sandia National Laboratones for the DG failure to start.

Nuclear Reputatory Commission. The Level 11/111 results are pre-NUREG/CR-4832 V03 P2: ANALYSIS OF THE LASALLE UNIT 2 sented in. associated reports desenbed in the Foreword. This NUCLEAR POWER PLANT: RISK METHODS INTEGRATION volume contains a summary descripton of the LaSalle plant, de-senbes the contents of the other nine volumes of this report AND EVALUATION PROGRAM (RMIEP). Internal Events Acci-dent Sequence QuantificatiortAppendices. PAYNE,A.C.;

and their relationships to each other, the relatonship of the La' DANIEL,S.L; WHITEHEAD D.W.; et al. Sandia National Labora-

.Salle program to other progrsms, a step-by-step summary de- tories. August 1992. 750pp. 9209220476. SAND 92-0537.

senption of the methodokigy and new techniques used to per- 63245'001 form the analysis, and presents tho integrated results obtained g,, NUREG/CR 4832,V03.P01 abstract.

by merging all of the accident sequence cut sets frorn the LOCA, transient, transient-induced LOCAs, and anbeipated acci- NUREG/CR-4832 V07: ANALYSIS OF THE LASALLE UNIT 2 NU, dents without scram accident sequences resulung from intomal CLEAR POWER PLANT: RISK METHODS INTEGRATION ANO initators with the cut sets from the fire, flood. and seismic anal-EVALUATION PROGRAM (RMIEP).Extemal Event Scoping yses accident acquences- Quantficahon. RAVINDRA,M.K.; BANON,H. NTS/SMA.- inc.

• Sandia National Laboratories. July 1992.161pp. 9208060219.

NUREG/CR-4832 V02: ANALYSIS OF THE LASALLE UNIT 2 NU- SAND 92 0537. 62654:060.

CLEAR POWER PLANT: RISK METHOOS INTEGRATION AND This report is a description of the scoping quantificaton study EVALUATION PROGRAM (RMIEP). Integrated Ouanbfication which selected the exWnal events to be included in the level And Uncertainty Ana!ysis. PAYNE,A.C.; SYPE T.T.: Hi PRA of the LaSalle County Nuclear Generating Station Unit WHITEHEAD.D.W.; et al. Sandia National Laboratones. Jufy 2. The study was performed by NTS/ Structural Mechansc:, As-1992. 669pp. 9208060243. SAND 92 0537 62648 001- sociates (SMA) for Sandia National Laboratones as part of the Thea volumo presents the methodology and results of the in- Level I anaiysis being pe' formed by the Risk Methods Integra-tegrated accident sequence analysis of the LaSalle Unit ll nu- ton and Evaluation Program (RMIEP); The r sthodology used is clear power plant. Integrated results are obtained by merging all desenbed in detad in a companion report, NUREG/CR-4839, in of the accident sequences' cut sets from the intemal and exter* this report, we desenbe the process for selochng the external' l- nal events analyses. The final dominant accident sequences are events, the sceening analysis, and the detailed bounding calcu-l determined and the integrated risk reduction, risk increase, and labons for those events not ehrninated in the screening analysis uncertainty importance measures are obtained. Also, an overall As a result of this anatysis, it was concluded that onry intemal ranking of the derninant cut sets was t btained. The total core flooding, intomal fire, and seismic events were potentW.y a, Jnifi-damage frequency at LaSalle from all events has a mean value cant at aSallec Detailed analyses were perform 1td for each of -

of t01E-04/R yr. with a 5th percentle of 5.34E 6/R- yr., a these and are reported in NUREG/CR4832, Volumes 10, 9, median value of 2.92E 05/R-yr., and a 95th percentde of 2.93E' and 8. respectively.

04/R yr. The dominant accident,35 4% of the core damage fre-quency, involves a loss of allinjection as a resuft of failures oc- NUREG/CR-4839: METHOOS FOR EXTERNAL EVENT SCREEN-cumng after a loss of offsite power. The dominant cut sets of ING OUANTIFICATION RISK METHOOS INTEGRATION AND -

this sequence represent a short-term station blackout type seci- EVALUATION PROGRAM (RMIEP) METHODS . DEVELOP-dont. The second most likely- sequence,17.2% of the core MENT. RAVINDRA,M.K.; BANON.H. NTS/SMA, Inc.

• Sandia damage frequency, is the result of a control room fire which is National Laboratories. Jufy 1992.126pp. 9208060210. SANDS 7-not suppressed and becomes large enough to require evacu- 7156. 62647;209.

abon of the contcol room. Auto actuation of the systems fails as in this report. the scoping quantfication procedures for exter-a result of the fire and the operators do not operate the remote nal events in probabWstic risk assessments of tiuclear power shutdown panel correctly due to the high stress Loss of all in- plants are desenbed. External event analysis in a PRA has jection occurs and short-term core damage resutts. three important goals- (1) The analysis should be complete in-t-

c;m A . m ,_-.m ,A.- ,,_..,.,n..,- m _ _ . ---- -.._-m _.,,- -._ --.,.~._ _ -.....a~ . - - - ~ ~ ~

. - - - ~ . - - _ . . . - - . - - - .- . .- . - - - .

e 8 Main Citations and Abstracts I

that all events are considered; (2) By following some selected into the near future and for supporting risk-based aging man- I screening cntena, the more signibcant events are identfred for 4

agement decisions _

detailed analysis; (3) The selected events are analyzed in depth by taking into account the unique' features of the events:

NUREG/CR-5416: TECHNICAL EVALUATION- OF GENERIC hazard, fragility of structures and equipment, external event iruti-ISSUE 113. OYNAMIC OUALIFICATION AND TESTING OF ated accident sequences, etc. Based on the above goals, exter. LARGE BORE HYDRAUUC: SNUBBERS. NITZEL M,Ea nal event analysts may be considered as a three-stage process: WAREAG. EGaG Idaho, Inc, PAGE,J.D. NRC No Detailed Af-Stage L identfcaton and Inabal Screening of Extemal Events; filiabon Gwen.' September 1992. 400pp. 9210130176. EGG-Stage 11: Bounding Analysis; Stage Ill: Detailed Risk Ana!ysis. In 2571 63461:003.

the prescat report, f4st, a revioW of pubbshed PRAs is gwen to focus on the signdicance and treatment of external events in This report summanzes the work performed by the Idaho Na-

}' futt-scope PRAs. Except for seismic, flooding, fire, and extreme bonal Engineenng Laboratory (INEL) for the Nuclear Regulatory wind events, the contnbutions of other external events to plant Commisson to resotve Generic issue 113. " Dynamic Qualifica-rrsk have been found to be neghgible. Second, scoping methods ton and Testing of Large Bore Hydraunc Snubbers (LBHSs)."

The report evaluates LBHS rehabihty and the need to improve for earternal events no' : overed in detail in the NRC's PRA Pro' that reliabihty. The INEL gathered and reviewed informatiori re .

cedures ses for trarqGum ortahonare provided accidents, extreme Forwmds thisand purpose,tornadoesbounding anaty',

garding snubber (including LBHS) numbers and uses, design, arreraft impacts, turbine rmssdes, and chemical release are de- environmental quahficabon, operating exponence, and the ef-scriberi fects of vanous snubber reduction programs. Limited qualitatue and quanhtatue analyses were performed regarding potential NUREG/CR-5305 V01: INTEGRATED RISK ASSESSMENT FOR LSHS single failures. A list of potentalimprovements to LBHS LASALLE UNIT 2 NUCLEAR POWER PLANT.Phenornenology rehability was generated and each item or; the list was evaluat.

And Rrsk Uncertainty Evaluation Program - (PRUEP). ed by probabilistic nsk and cost / benefit analyses. Eleven rec.

BROWN T.Da PAYNEACa MILLER.LA ; et al Sandia National ommendations were made; five applicable to existing and future Laboratories. August 1991385pp. 9209220410, SAND 90 2765. plants, fue apphcable only to future plants, and one for further -

6'1243.001. " single-failure" research.

A Level 111 probabdistic nsk assessment (PRA) was performed for the LaSane Unit 2 nuclear power plant. The objectwa of this NUREG/CR 5443: CORE-CONCRETE INTERACTIONS USING study was to provide an estimate of the nsk to the offsite popu-tation donng full power operabon of the plant and to include a MOLTEN URANIA WITH ZlRCONIUM ON A LIMESTONE CON-CRETE BASEMAT.The SURC-1 Expenment. COPUS.E.R.;

charactenzabon of the uncertaintes in the calculated nsk BROCKMANN,J Ea et al, Sandia National Laboratories.

values. Uncertainties were included in the accident frequency BLOSE.R.E. Ktech Corp. September 1992.275pp.9210150151.

analysis, accident progression analysis, and the source term SAND 90-0087. 63525:039.

analysis. Only weather uncertainties were included in the conse-quence analysis. The nsk eshmates presented in this report in- An inductvely heated evpenment, SURC 1, was executed as part of the integral Core-Concrete Interactions Expenments Pro-ciude cetnbutions from both internal and external initiators grarrt Tho purpose of this expenrnent was to measure and -

The offsite nsk to the public due to the operaton of LaSalle County Station is relatweiy low, especialty with respect to the assess the vanety of source terms produced dunng core debns/

NRC safety goals. The mean indwidual earty fatality risk within 1 concrete interactons. These source terms include thermal mile is 1,1E-10/R-yr which is more inan three orders of magni- energy released to both the reactor basemat and the contain-tude below the safety goat. S+mdarty, the mean indwidual latent ment environmen*,. as well as flammable gas, condensaDie -

cancer fatanty nsk is 8.5E 09/R yr which is slight!y more than vapor and toxic or radioactuo aerosols generated during the two orders of magnitude below the safety goal. In fact, the course of a severe reactor accident. The SURC-1 experiment entire uncertainty distnbutens for these two nsk measures lie used 204 kg of prototypic UO(2)-ZrO(2) materials to study the below the safety goals. The mean va;ues for earty fatahty nsk interactions betwee1 core debris and a limestone basement.

and for latent cancer fatahty nsk are L2E-08/R yr and 0.25/R. The experiment eroded 27 cm of concrete dunng 130 minutes A res m twey of sustained interaction at temperatures which ranged from 2,650 to 2,200 K. Comprehensue gas flow rates, gas composi-NUREG/CR-5378: AGING DATA ANALGIS AND RISK ASSESS- tons, and aerosol release rates from the interaction were also MENT--DEVELOPMENT AND DEMONSTRATION STUDY. measured.

WOLFORD.A.J. DNY Technica. ATWOOO.C.La ROESENER.W.S4 et af. EG&G Idaho, Inc. August 1992. 253pp. NUREG/CR-5564: CORE-CONCRETE INTERACTIONS USING 9209240315. EGG-2567. 63273.00t This work develops and demonstrates a probabihstic nsk as- MOLTEN UO(2) . WITH ZlRCONIUM ON A BASALTIC BASEMAT.The SURC-2 Expenment.- COPUS,E.R.,

. sessment (PRA) approach to assess the effect of aging and BROCKMANN,J E.: et at Sand.a National Laboratories.

degradaten of actue components on plant nsk, The work: (a) BLOSE.R.E. Ktech Corp. ' August 1992. 297pp. 9209220462.

develops a way to identify and quantify agedependent failure SAND 901022. 63237:001-l rates of actae components, and to incorporate them into PRA; . An inductivety heated experiment, SURC-2, was executed as -

(b) demonstrates the approach by applying it to a fluid mechani-

'part of the Integral Core-Concrete Interactions Experiments Pro cat system, using the key elements of a NUREG-1150 PRA; and gram. The purpose of this expenment was to rneasure and 1

(c) presents it as a step-by-step approach, to be used for evalu-assess the variety of source terms produced curing core debris /

attng the risk significance of aging phenomena in systems of in-concrete interactions. These ' source terms include thermal' terest. The approach uses statist. cal tests to detect increasing energy released to both the reactor basemat and the contain-failure rates and for testing data-pooling assumptions' and model adequacy. The component failure rates we assumed to ment environment, as well as flammable gas, mdensable vapor and toxic or radioactue aerosols generate.J during the change over time, with several forms used to model the age de- course of a severe reactor accident. The SURC-2 expenment pendence - exponential, Weibuit, and hnear. Confidence inter-vals for the age-dependent failure rates are found and used to used 200 kg of prototypic UO(2FZrom matenals to study the -

interactions between core debris and a basaltic basement, The develop tnputs to a PRA modelin order to determine the pitnt expenmer,t eroded 35 cm of concrete dunng 160 minutes of core damage frequency The approach was used with plant-spe-cific data, obtarned from rneintenance work requests for the sustained interaction at temperatures which ranged from 2700 auxiliary feedwater system of an older pressurried water reac- to 2200 K Comprehensive gas flow rates, gas compositions, tor. The approach can be used for extrapolating present trends and acrosd release rates from the interaction were also meas-ureci i

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d -

Maln Citations and Abstracts 9 NUREG/CR 5673 V03:  : TRAC-PF1/ MOD 2 CODE-NUREG/CR 5587: . APPROACHES- FOR ~ AGE-DEPENDENT -

MANUAL. Programmer's Guide GUFFEE LA.- Science Applica.

PROOABILISTIC SAFETY ASSESSMENTS WITH EMPHASIS tions internatonal Corp. (formerty Science Applicatons, inc1 ON PRIORITIZATION AND - - SENSITIVITY STUDIES WOODRUFF,S B.; STEINKE.R.G.; et at Los Alamos Natonal VESELY,W E. Science Applications international Corp. (formerly Laboratory. July 1992f 34tpp. 9206060249. LA-12031 M.

Science Appl +catons. Inc.), August 1992.164pp. 0209220333.

62653:00f.

SAIC-92/113 /. 63240.099. The Transient Reactor Analysis Code (TRAC) was developed Approaches are described for incorporating component aging to provide best-estimate predictions of postulated accidents of rehabihty models into a prooabd.stic safety assessment (PSA), lig%watu reactors. The TRAC-PF1/ MOD 2 program providen or probabdistic risk assessment (PRA). of a nuclear power plant _ this capabihty for pressurized water reactors and for many ther-These approaches and procedures are described from a techni- ma -hydraulac test facilibes. The code features either a one-or a cal standpoint and are not to be interpreted as having any regu- three-d.mensonal treatment of the pressure vessel and its as-latory iniphcations. Component a9i 09 failure rate models and sociated internals, a two- fluid nonequdibdum hydrodynamics -

test and maintenance aging control models are presented for model with a noncondensable gas field and solute tracking,

' utilizatton. Different approaches for carrying out the aging eval- flow regime-dependent Constitutve equation treatment, opicnal untions are given. Demons %ons are given involving pnontmng reflood tracking capabAty for bottom-flood and falling-film aging contnbutors, evaluabng maintenance effectiveness, carry- quench fronts, and consistent treatment of entire accident se-ing out tme dmendent evaluations, and carrying out uncertainty quences, including the generation of consistent initial condi-and sensitiv&; nalyses of aging effects- tions. This manua: is toe third volume of a four-volume set of -

documentaten on TRAC-PF1/ MOD 2. This guide was developed NUREG/CR 5646: PIPING SYSTEM RESPONSE DURING HIGH- to assist the TRAC programmer and contains information M the LEVEL SIMULATED SEISMIC TESTS AT THE HEISSDAMP' TRAC code and data structure, the TRAC calculational se-FREAKTOR FACluTY (SHAM TEST FACILITY). STEELE.R_; quence, memory management, and various machine configura-f'

' NITZEL.M E. EG&G Idaho. Inc. July 1992. 231pp. 9208260229- bons supported by TRAC.

EGG-2655. 62876:155.

The SHAM seismic research program studed the effects of NUREG/CR 5685: SEALING PERFORMANCE OF BENTONtTE AND BENTON!TE/ CRUSHED ROCK BOREHOLE PLUGS.

increasing levels of seismic excitaten on a full-scale. in situ nu. OUYANG.S.; DAEMEN.J.J.K. Arizona Univ, of Tucson, AZ. July clear ppng system containing a naturafty aged United States 1992. 340pp 9208260235. 62877.026.

(U.S ) 84n. motor operated gate valve. The program was con _ This study inciudes a systematic investigation of the sealing ducted by Kerntnmchungs2entrum Karlsruhe at the Heissdamp- performanco of bentonite and bentonite / crushed rock plugs.

treaktor near Frankfurt, Germany.' Participants included the Amencan Cottosd C/S granular bentonne and crushed Apache United States, Garmany, and England. Fifty one expenments Leap tuft have been mixed to prepare samples, Bentonite were conducted, with the pping supporied by sm different piping weight percent and crushed tuff gradation are the major varia.

support systems, inclu6ng a typcal sttf U S. piping suppon bles studied. Hv)h injection pressure flow tests, polyaxial flow system of snubbers and ngid struts. This report specifically ad' tests, high temperature flow tests, and piping tests have been dresses the tests conducted with the U.S. system, The piping pertarmed. A composition to yield a permeabdity lower than 5 x system withstood large d:splacements caused by overload 10( 8) cm/s would have at least 25% bentorate by weight mixed snubber fadures and local piping strains. Although some limit with well-graded crushed rock Hydraulic propmties of the mix-switch chatter was observed, the motor operator and vaive ture plugs may be highly anisotropic if significant particle segre-funcboned smoothly throughout the tests. The results indicate gation occurs during sample installation and compactort Tem-that sufficient safety marg!ns exist when commonly accepted perature has no significant effect on sealing performance from desrgn methods are apphed and that pipeng systems wd6 IAety room temperature to 60 degrees C. Piping damage is small if maintain their pressure boundary in tne presence of severe the hydraulic gradient does not exceed 120 and 280 for sam-loading and the loss of multiple supports. pies with a bentonite content of 25 and 35%, respectively. Tne hydraucc gradents above which flow of bentonite may take NUREG/CR 5673 V02: TRAC-PFt/ MOD 2 CODE MANUALUser's place are deemed entical. Bentonite occupancy percentage and Guide. SCHNURR,N.M; STEINKE,R.G.; MARTINEZ,V.; et a!. water content at saturation are two major parameters for plug.

l design. A model is developed for predcting the permeabsty irt --

Los Alamos National Laboratory . July 1992. 883pp.

clays. A piping rnodel, based on plastic flow theory, permits esti- ,

9208240311 LA-12031 M 62867:001. '

The Transient Reactor Analysis Code (TRAC) was developed mat ng the critical hydrauhc gra$ents at which flow of bentonite takes placecThe model can eo be used to define the maxi-

to provide advanced best. estimate predictons of postufated ac. mum allowable pore diameter i a protective fdter layer.

cidents in pressunzed light-water reactorsJ The code features either a one- or a three dimensional treatment of the pressure NUREG/CR-5687: BOREHOLE STABILITY IN JDENSELY vessel and its associated internals, a two-fluid nonequihbrium WELDED TUFFS. FUENKAJORN,K.; DAEMEN.J,J K, Arizona, hydrodynamics model with a noncondensable gas field and Univ. of, Tucson, AZ, July 1992. 71pp. 9208250287. 62871:158.

solute traciung, flow 4egime-dependent consttutive equaton Fadure of host rock at seal locations may altow bypasa flow treatment, optional reflood tracking capabihty for bottom-flood around seals. This report presents an snvestigation of compres-and falling film quench fronts, and consistent treatment of entire sive failure around boreholes in densely welded Apache Leap accident sequences including the generatton of consistent inittal tuff.' Triaxial and polyaxial tests have been performed on cylin -

cond tions. In add: tion to the compononts contained in previous dets and blocks containing coaxial cored holes. Test hole diam- .-

TRAC versions, TRAC-PFI/ MOD 2 includos a heat-structure eters are 14 mm for tnaxial tesung and 25.4 mm for biaxial test-

' component that allows the user to accurately modet complicat . ino. To induce breakouts requires stresses that exceed elasti-ed geometnes. An improved reflood model based on mechanis. cally calculated boundary stresses equal to the uniaxial com.

tic and defensible models has been added The now code also pressive strength. Failure pattems are influenced by heteroge-contains improved constitutve models and additions and refine- neity:- soft inclusions fail first. Such failures remain. localized.

ments for several components. This Guide deacnbes the compo. The stronger surrounding matnx maintams hole stahdity. An elastic analysis of hole stabihty in welded tuff may provide a sig-3 nents and control systems used in TRAC and gives detaded in.

nitant safety margin. This conclusion needs to be qualified: all formation the user needs to prepare an input dock and carry out expenments have been conducted on small diameter boreholes.

simulations using TRAC PFI/ MOD 2. It would be desirable to conduct borehole stabdity experiments

, ,wn.n.,,,.- ,n,n.-.-, _ n ,n,..,~,,._._.---,...-,.-,..,..n..-~, . ~ ., . ~. - - . ,,- - , ,

' 10 - Main Citations and Abstracts on larger holes Of particular importance may be the influence elongation, modutus. And density) were more effective than of flow layers in tuff on borehole stability. Effects of fiow layers electrical traesurements for monitoring age-related degradation. l

- have been minernzed by prepanng all samples normal to the  !

flow layers. Also desirable would be an investigation of the influ- NUREGM 5779 V01: AGING OF-NON.POWERICYCLE HEAT ence of environmental condibons, especialty temperaturo end EXCHANGERS USED -IN NUCLEAR POWER- '

moisture content, and of the strength under sustained long term PLANTS Operahng Experience And Fadure identification.

loading. A more comprehenshte analysis of the results should MOYERS,J C. Oak Ridge Nabonal Laboratory July 1992.70pp' I be performed ne ng evaluation of recent theorehca! models 9208250292. ORNL 6687. 62871:090 a d b pmM d h W l re!ated degradation of non-power-cycle heat exchangers used NUREG/CR 5700: AGING ASSESSMENT OF REACTOR IN- in nuclear power plants. The assessment was sponsored by the STRUMENTATION AND PROTECTION SYSTEM U.S Nuclear Regulatory Commission's Nuclear Plant Aging Re-COMPONENTS. Aging-Related Operating Experiences- search Program. Heat exchanger design characteristics and ap, GEHL.A C.; HAGEN.E.W. Oak Ridge Natsonal Laboratory. July plicatons in the plants are desenbed and stressors leading to 1992.154pp.9208260222. ORNL/TM 11806. 62876 001, degradation are identified Opotabng experience, as identified A study of the aging-related operating expenences throughout from nuclear industry data bases, is reviewed and failure types a five year penod (19841988) of six generic instrumentaten and causes are summarized. Regulatory requirements for in-modules Ondicators, sensors, controllers, transmitters, annunci- specton and tesbng, with a bnef discussion of Industry prac-ators, and recorder $f was performed as a part of the USNRC xes in this area, are presented.

Nuclear Plant Aging Research Program. The effects of aging from operahonal and environmental stressors were character, NUREG/CR-5787 V01: TIMING ANALYSIS OF FWR FUEL PIN ized from resutts depicted in Licensee Event Reports (LERs). FAILURES. Final Report. Main . Text And Appendices A-J.

The data are graphically displayed as frequency of events per JONES K.R.; WADE,N L; KATSMA.K.R.; et al. EG&G Idaho, plant year for operating plant ages from 1 to 28 years to deter. Inc. September 1902. 350pp. 9210050053. EGG-2657.

mine aging 4 elated failure trend patterns Of the six modules 63382:205, studied, indicators, sensors, and controllers account for the bulk Research has been conducted to develop and demonstrate a (83%) of aging-related failures, infant mortahty a;) pears to be methodology for calculaton of the tsmo interval between receipt the dominant failure mode for most 8&C module categones. Of of containment isolaton signals and the first fuel oin failure for the LERs issued dunng 1984 1988 which dealt with malfunc. loss- of-coolant accidents. Demonstraton calculatons were per-tons of the six instrumentat>on and control rnodules studied, formed for a Babcock and Wilcox (B&W) design (Oconee) and a 28% were found to be aging- related (other studies show a Westrnghouse (Yl) four- loop design (Seabrook). Sensitivity range of 25-50%). studies assessed the impacts of fuel pin bumup, axial peaking factor, break size, emergency core coohng system availability, NUREG/CR 5758 V02: FITNESS FOR DUTY IN THE NUCLEAR and main coolant pump tnp on these times. The analysis was POVvER INDUSTRY. Annual Summary Of Program Performance perictmed using FRAPCON-2 and FRAP T6 for the calculation Reports.CY 1991. MURPHY,S.; FLEMING,T4 WESTRA.C.; et al. of steady-state and transient fuel behavior and SCDAP/

Battelle Human Affairs Research Canters. August 1992, 90pp. RELAPS/ MOD 3 and TRAC-PF1/ MODI for the calculation of 9209240238. PNL 7736. 63279 278. transient thermal.hydrauhc condihons in the reactor system.

, This report summanzes the data from the semiannual reports Using SCDAP/RELAPS/MOC1 and TRAC-PF1/ MOD 1 thermal-on fitness-for duty programs subnutted to the NRC by 54 utihties hydrauhc data, the shortest a ne intervals cal..ated between for two reporting ponods: January 1,1991 to June 30,1991 in:tiation of containment isolation and fuel pin fadure are, re-and from July 1,1991, to December 31,1991. Dunng CY 1991, spectively,10.4 and 10.3 seconds for Oconee and 19.1 and hcensees reported that they conducted 262,597 tests for the 29.1 seconds for Seatxook. These intervals are for a double-presence of Jiegal drugs and alcohot Of these tests.1,721 ended, offset-shear, cold leg break usrng ine technical specifi.

L66%) were positive. Positive test results varied by category of cation maximum peaking factor apphed to fuel with maximum test and category of worker. The maiority of positive test results design burnup. Using peaking factors commensurate with actual

- (983) were obtained through pre-access test!ng. Of tests con"- burnups would result in longer intervals for botn reactor designs.

ducted on workers having access to the protected area, there were 509 positive tests from random testing, and 167 positive NUREG/CR 5787 V02: TIMING ANALYSIS OF PWR FUEL PIN tests from for cause testing. Followup testing of workers who FAILURES. Final Report. Appendices K L. -JONES.K.RJ had previously tested positive resulted in 62 positive tests. Posi- WADE.N.L; KATSMA.K.R.; et at. EG&G Idaho, Inc. September tive test reotts also vaned by category of worker, Overall, 1992. 400pp. 9210050062. EGG 2657. 63383:105.

short terth and long-term contractor personnel had the highest See NUREGICR-5787,V01 abstract.

rates of positive tests Licensee employees had lower rates of posite test results' NUREG/CR 5790: RISK EVALUATION FOR A B&W PRESSUR.

IZED WATER REACTOR. EFFECTS OF FIRE PROTECTION NUREG/CR-5772 V01: AGING. CONDITION MONITORING, AND SYSTEM ACTUATION ON SAFETY RELATED LOSS-OF-COOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E EQUIPMENT. Evaluation Of Genene issue 57. LAMBRIGHT,J.

ELECTRIQt CABLES Crosahnked Polyolefin Cables. Sandia National Laboratories. LYNCH,J.: ROSS.S.; et at Sci-JACOBUS #.J. Sandia National Laboratones. August 1992, ence & Engineenng Associates, Inc. September 1992. 250pp.

200pp. 2209240244. SAND 91-1766/1. 63263 001. 9210150130. SAND 91-1535. 63525:317.

This report describes the results of aging, condition monitor. Nuclear- power plants have experianced inadvertent actu-eng, and accident testing of crosshnked polyolefin (XLPO) ations of fire protection systems (FPSs) under conditions for cables. Three sets of cables were aged for up to 9 months which these systems were not intended to actuate, and also under simultaneous 5ermal- ( = 100'C) and radiation have experienced advertent actuations with the presence of a

' ( = 010 kGy/hr) conditions. A sequential accident consisting of high fire. These actuations have often damaged plant equipment. A dose rate irradiation (= 6 kGy/hr) and high temperature review of the impact of past occurrences of both types of such steam followed the a0!ng The test results indicate that most events on nuclear power plant safety has been performed. TNr.

properly installed XLPO cables should be able to survive an teen different scenarios leading to actuation of fire protection accident after 60 years for total aging doses up to 400 kGy systems due to a variety of causes were ident.fied. These sce-and for moderate ambient temperatures on the order of narios ranged trem inadvertent actuation caused by human 50 55* C (potentially higher or lower, depending on rnatorial spe. error to hardware Wure, and includes seismic root causes and cific activation energies). Mechan'il measurements (pnmarily sersmic/ fire interaction. A quantification of these thirteen sce-L

[;

1 Main Citations and Abstrads 11 I

narcs, where applicable, was periormed on a Babcock and tion to a number of other f actors, has caused the Nuclear Regu-Wilcox (B&W) Pressunzed Water Reactor Oowered luop design) latory Comnussion (NRC) to consider conservative assumptons This report estimates the contnbuton of f PS actuatons to core rsgarding fuel behavior. For this purpose, the concept termed damage frequency and to nsk. " weak fuel" has been proposed on an interim basis "Wuak f# m a penalty imposed on consequenet analyses whereby NUREG/CR 5793: A COMPARISON OF AN ALYSIS METHOD, the fuel is assumed to respond less favorably to environmental OLOGIES FOR PREDICTING CLEAVAGE ARREST OF A DEEP condions than predicted by behavioral models The rationale CRACK IN A REACTOR PRESSURE VESSEL SUBJECTED TO for adopting this penatry. as well as conditions that would permit PRESSURIZLD-THERMAL-SHOCK LOADING CONDITIONS. its reduction or ehmination, are osamined in this report The KEENEY-WALKER. BASS.B R. Oak Rdge National Laboratory eva alcn od@s an examaton d pod M mandam-September 1992 31pp. 9209290362. ORNL/TM 11969 ing defects, quality-control procedures for defect detecten, and 63336 001 the mechanisms by which fuel defects may lead to failure.

Several calculational procedures are compared for predicting cleavage arrest of a deep crack in the wall of a prototypical re- NUREG/CR-$819: PROBABILITY AND CONSEQUENCES OF actor pressure vessrA (RPV) subjected to pressanied thermal- RAPID DORON DILUTION IN A PWR.A Scoping Stur;y.

shock (PTS) types of loading cond; tons Three procedures ex- DIAMOND.D.J.; KOHUT,P.; NOURBAKHSH,H.; et al. Brookha-amined in this study used the following models. (1) a static von National Laboratory. June 1992. 98pp 9207270295. DNL-finite-element model (full bending). (2) a radially constrained NUREG 52313,62512:115.

static model, and C) a thermoelastic dynarnic finite-element This report documents the results of a scoping study of rapsd model A PTS transient loading cond: tion was selected that pro- diluton events in pressunzed water reactors. It reviews the sub-duced a deep arrest of an axially onented, initialty shallow crack ject n broad terms and focuses on one event of most interest.

according to calculational resutts obtained from the static (full. This event Could occuf dunng a restart if there is a loss.cf-off-bending) model. Results from thn two stahc models were com- site power when the reactor is being deborated. If the volume pared with those generated from the detailed thermoelastic dy- control tank is filled with water at a low boron concentraton namic tinite-element analysis. The dynamic analyses modeled then a slug of this water could accumulate in the lower plenurh.

cleavage-cnick propagabon using a nod + release technique and This would be the resoft of the tnp of the reactor coolant pow 6r.

apphcation- and Generation-mode methodologies Compareons The concern is that this diluted slug will rapidly enter the core preser,ted here indicate that the degree to which dynarruc solu- after a reactor Coolant pump ts restar1ed and this could cause a tions can be approximated by static models is highly dependent power excursion leading to fuel damage. This problem was on several factors, including the malmial dynamic fracture stud ed probabilist cally for three plants and the irnportant curves and the propensity for cleavage reinet;ation of the arrest' design features that affect the core damage frequercy were ed crack under PTS loading conditions Additional work is re- identified. This anatysis was augmented by an analysis oi the quired to develop and vahdate a satisfactory dynamic fracture mmng of the dduted water with the borated water already toughness model apphcable to postcleavage arrest conditons in present in the vessel. The mming was found to be significant so an RPV' that neglect of this mechanism in the probabihstic analysis leads NUREG/CR 5804: REPOSITORY OPERATIONAL CRITERIA to very conservatrve results. Neutronic calculatons for one plant ANALYSIS. HAGEMAN.J P.; CHOWDHURY,A H Center for Nu- werc camed out to understano the effect of nuclear design on clear Waste Regulatory Analyss August 1992. 391pp the consequences of the event 9209240189 CNWRA 91014 63270.001. NUREG/CR-5840; FASTGRASS- A MECHANISTIC MODEL FOR The obiectue of the " Repository Operatona! Critena (ROC) THE PREDICTION OF XE, I, CS, TE, BA, AND SR RELEASE ,

Feasitxtity Studies" (or ROC task) was to conduct comprenen.

sue and integrated analyses of repository design, constructon, FROM NUCLEAR FUEL UNDER NORMAL AND SEVERE-AC-CIDENT CONDITIONS User's Guide For Mainframe, Worksta-and operations critona in 10 CFR Part 60 regulatons, consider. And Personal Computer Apphcations. REST J ;

tion, ing the interfaces and impacts of any potential changes to ZAWADZKI,S.A. Argonne National Laboratory. September 1992.

those regulatons The stuc*y addresses regu!atory criteria relat.

ed to the proclosure aspects of the geologic repository The 179pp. 9209240206. ANL 92/3 6327t:032.

The pnmary physical / chemical models that form the bests of study task developed regulatory concepts or potential repository the FASTGRASS mechaniste compuw modet fc, calculating operational cnteria (PROC) based on analysis of a repostory's safety funct;ons and other regulahons for similar facilities. These hsson-product release from nuclear fuel are desenbed. Calcu-lated results are compared with test data, and 'he maior mecha4 regulatory concepts or PROC were used as a basis to assess nisms affect,g the transport of fission products during steady-the sufhciency and adequacy of the current entena in 10 CFR state and accident conditions are kjentihed Part 60 Where the regulatory concepts were the same as cur.

rent operahonal critena, these entena were referenced.1he op- NUREG/CR 5849 DRF FC: MANUAL FOR CONDUCTING RADI-erabor,s entena referenced or the PROC develooed are given in this report Detailed analyses used to develop the regulatory OLOGICAL SURVEYS IN SUPPORT OF LICENSE TERMINATION Draft Report For Comment. BERGERJD. Oak concepts and any necessary PROC for those regulations that Ridge Associated Univers2hes. June 1992 207pp 9208060313.

may require a minor change are a!so presented The results of ORAU 92/Cb7 2647 00 the ROC task showed a need for further analysts and possible magor rue change related to the design bees of a gentogic re' cal surveys dunng decommissioning. to demonstrate that residu-pository operatons area, siting. and radiological emergency al radcactive matenal satisfies entena estabhshed by the U S.

""'"9 Nuclear Regulatory Commission (NRC) for terminat>on of a le fiUREG/CR 5810: EVALUATION OF MHTGR FUEL RELIABIL- cense. The manual denenbes procedures for design ar# corv ITY. WICHNER,R.P. Oak RK1ge Natonal Laboret iry. duct of surveys in a manner which will provide a high degree of DARTHOLD,W.P. Barthold & Associates, Inc. July 1932.1 epp- assurance that NRC guidelines and conuttons have been sat 2s-9208240298. ORNL/TM 12014. 62860120 hed. The manual also describes methods for documer' ting the Modular Higt> Temperature Gas-Cooled Reactor (MHTGR) survey hndings in a final report to the NRC. This manual up-concepts mat house tne reactor vesset in a tight but unsealed dates information contarned in NUREG/CR 2082, "Monitonng reacto malding place heightened importance on the reliabihty of for Cornpitance With Decommissioning Terminaton Survey Cth the fues particle coatings as hssbon pr%ct bamers Though ac- teria. tORNL 1981)." ft incorporates statistical approaches to cident consmem c analym 'oeN to show favorable re- survey design and data anterpretation used by the Environmen-

} suits, the inci : 1 dependence on one type of bamer, in adds tal Protecten Agency for evaluation of hazardous materials i

12 Main Citations and Abstracts sites under Superfund (CERCLA). Quahty essurance is empha- wotten for the IBM personal computers or their compatibles to sued ttroughout evaluate the rehabihty of wolds in nucioar power plant piping syst ma pc-PRAISE was adapted from tm PRAISE computer NUREQ/CR 5850 UNIVERSAL TREATMENT OF PLUMCS AND code which was onginally daveloped for CDC 7600 computort STRESSES FOR PRESSURIZED THEhJAL SHOCK EVALUA.

TIONS THEOF ANOUS,T G ; ANGEUNI.S : YAN.H Cahfornia, in 1981 by Lawrence Livermore National Laborat ry under fund.

Univ, cf, Santa Darbara, CA. June 1992 96pp. 9207270318 ing from the U.S, Nuclear Regulatory Commisson, and has been considerably expanded and updated over the years 62508.146.

The thermal field in a reactor vessel downcomer are resulting NUREG/CR 5867: GRADIENT STUDY OF A LARGE WELD JOIN.

thennal/ stress response in the adjacont reactor vessel wall ING TWO FORGED A 508 SHELLS OF THE MIDLAND REAC-during fugh-pressure safety intecton are examined, especially TOR VESSLL IRWIN,0 R.; ZHANGXJ. Maryland, Univ. of. Cob with togard to departares from one-dimensional behavior. S"* lege park, MD.

• Oak Ridge Natonal Laboratory. June 1992.

lanty solut ons for the strabfication (in the cold leg) that creates 19"p LO7270284. ORNLSUB79777810. 32508.287.

the downcomer plumes, and scahng considerahons for the the

The low-carton welds (WF67 and WF.m M i.ie slab exam-mM conduction and stress fields in the vessel wall are devel' innd contained no abnormahbes that would ndicate fracture be-oped to provide generah7od entena for the adequacy of th" havior different from that observed in bulk matenal fracture toe-dimensional treatment' tests The A 508 matonal in the HAZ regon, very close to the NUREG/CR-58L9: MODELING THE INFLUENCE OF IRRADIA, welds, contains small (3mm) regions ad iacent to each L^yer of TION TEMPERATURE AND DISPLACEMENT RATE ON RADt. weld runs where grain coarsening and hardness elevation sug- _

ATION-INDUCCD HARDENING IN FERRITIC STEELS. gest reduction of cleavage initiaton toughness The degree of STOLLER.R E. Oak Ridge Natonal Laboratory. Juty 1992 40pp severity is largest where thi local region coincides with a local 9208250219 ORNL/TM 12073 62884.152 etovahon of cartw1e density in the A 508 matenal. The A 500 The influence of irradiation temperature and dispuement HAZ regon adi a t to the topmost weld run may be the region rate have boon investigated using a model based on the reac. most hke'y to P ,1 cleavage-fiacture initiahon because of its tion rate theory descnption of radiabon damage This theory locat on: close to free surface, smal! cracks, and the HAZ was developed primanty for the invesSgaton of Natively high. tegion beneath the uladding 11 was noted that the small cracks temperature, high dose radiation effects such as void swelling under the cladding have the appearance of pnor austenite grain and irradiaton croop Before apptystig that theory to the much boundary separabons that connect to austerute grain bound-lower temperatu.a and dose regimes charactensbc of light anes in the cladding The mtreme hardness ci a narrow layer of water reactor pressare vessels and support structures, it is noc- cladding at the fusion boundary may be of interest in further essary to examine the assumptions made in formulating the Studies of claddng toughness theory. The maior simpidying assumption that has commenty

< been made is that the interstitial and vacancy concentrabons NUREG/CR 5868: DEVELOPMENT OF POSITION SENSITIVE reach a quasi 4teady state condition rapidly enough that the PROPORT ONAL COUNTERS FOR HOT PARitCt.E DETEC.

steady state concentrabons can be used in calculahng tho ob- TION IN LAUNDRY AND PORTAL MONITORS SHONKA.J J:

servable radiation oriects. The resutts presented here indicate SCHWAHN.S 0 ; DENNETT,T E4 et al Shonka Research Asso-that the assumption of steady state point defect concentrabons crates, Inc. Septemt.or 1992. 38pp 0210150137, SRA-9201.

is not vahd for temperatures much below the light water reactor 225 00t pressure vessel operahng temperature of about 288 degrees C This report summanzes research which demonstratew he use At lower temperatures, the time required for the point defect Of P05' tion sensitwo proportional counters in contaminat on concernrabons to reach steady state can exceed an operahng monibnng syes. Both laundry moriitoring and portal monitor-reactor's lif etime. Even at 288 degrnes C, the point deloct tran. ing systems vere deployed The laundry monitor was deployed sient time can be long enough to influence the interpretahon of at a nuclear power plant where it was used to rmnstor clothing arradation empenments done in matenals test reactors st accel. dunng an outage Posthon sensitwe propcrtional counter based ersted darnage rates. Dased on the insights obtained with the contas.nnahon monitonn; %Wms were shown to have signih-simple modets of port defect evolution, a more dotatted model cant advanta F over systems using conventional proportional was devatoped N incorporates an exphc3 descriphon of pamt r gunters. These advantages include the abihty to directly meas-defect clustenng These clusters are pote7Dah ra,wnsible for ure the atea and quantty of contammaton. This capability per-the traction of the radinhon-induced hardening that ts attnbuted nWs identificahon of hot particles These systems are a!so capa-tc the so-called "matnx defect " The mc$el consders both in- ble of sett eckbeatroit via internal check sources Systems de-terst t:al god vacancy clustering The former are treated as pioyed with this technology should benef t from reduced com.

Frank tops wble the latter are treated as microvoids The point pleuty, e st, and maintenanco The inherent reductioh of back-detect clears can be formed etther directly in the displacement ground that occurs when the counter is electronically divided cascade or by ddfusive encounters betwen free point defects. ento numerous detectors perm 4ts operation in high background The results of molecutar dynamics simutaten studies aro used radiation fmids and improves detechon limits over convenhonal to provide guidance for the clusterrg parameters The harden, technology. i ing due to point defect clusters was calculated using a simple dislocation bamer model. The results indicate that both interst -

NUREG/CH-5872: ORNOZL A l'lfdEuteMENT MESH GENER.

ATOR I-On NOZZLE CYLINDER INTERSECTIONS CONTAIN-tial and vacancy clusters can give nse to signdicant hardening ING INNER CORNER CRACKS KEENEY WALKER; BASS.B R The relative importance of each cluster type is shown to be a function of inadiation temperature and displacement rate' Ook Ridge Nabonal Laboratory. September 1992. 42pp 92100S0048 ORNL/TM-11049. 63382164 NUREG/CR 5864: THEORETICAL AND (MR'S MANUAL FOR This report desenbes the ORNOZL finite-element mesh gen-PC-PRAISE A Probabilishc Fracture Mechanics Computer Code erator program for computational hacture mechanics analysis For Piping Rehabihty Analysis. HARRIS.D O.; DEDHIA,0.D. Fail- The program automaticsity generates a three dimensional (3 0) ure Andysis Associates, Inc. LU.S C. Lawrence Lrvermoro Na- finite-element model for four ddferent geometries of a corner honal Laboratory July 1992 352pp 9208060116 UCRL-ID- crack in a nozzle cytmder intersecuan. ORNOZL generates a 109798. 62f30 00 t core of special wedge or collapsed pnsm elements at the Crack This document consolidates and updates the earher reports front to introduce the appropriate stress singutanty at the crack which prov+de the theoretical background as well as information tip Reguiar 20-noded isoparametnc bnck elements are used needed for the emocution of the computy code pc PRAISE pc- away from the cracii front in the modehng Also, an option is PRAISE is a probabibstic fracture mechanics computer code included that allows for on embedded or penotrat ng crack in

i Main Citations and Abstracts 13 clad matonals As few as fwe input cards are required to exe- stra:nt loss but less than en the shallow crack test specimens.

cute the progiam. ORNOZL is part of a three-program system, Considenng the RPV in terms of Jantegral and 0-stress sug-ORNO7L ADINA- ORVIRT, which addresses linear or nonlinear gests there may be a larger margin of safety than would be fracture in 2 or 3-D croca geometnes. ORNOZL creates files found using the J. integral alone ' hermabshock data, which containing Todal point coordinates and element connectivit.es were generated using cylindrical vessels under thermal shock that have formats compatible with the ADINA structural analysis loading, show no significant increase in toughness even for program. ORVIRT is a post processor to ADINA and employs a shallow-flaw depths The thermal shock data seem to indicate virtual crack ortonsion technique to compute energy release two offsetting effects: a shallow tlaw effect, which increases rates at specihed positons along the crack tront. toughness, and an out of-plane (b: axial) stress effect, which de-NUREG/CR 5880: NONISOTHERMAL HYDROLOGIC 1RANS- creases toughness. Additional work is necessary to resolve out-PORT EXPERIMENTAL PLAN RASMUSSEN,T.C ; EVANS D.D standing issues for applytng shallow-crack data to an RPV and Anzona. Unw. of, Tucson, AZ. September 1992. 53pp vahdating the J.O technique for treture evaluations.

9209240196 63275 251' NUREG/CR-5891: ACCELERATt;D IRRADIATION TEST OF A field heater expenmental plan as presented for investgating hydrologs tranvo't processes in unsaturated fractured rock ro- GUNDREMMINGEN REACTOR VESSEL TREPAN MATERIAL lated to the disposaf of high level raioactivu waste (HLW) in an H AWTHORNE,J R. Matonals Engineenng Associates, Inc.

August 1992. 78pp. 92092402B4 MEA 2468. 63276.001, undergrourW Npository. The expenmental plan providos a meth' odology for obtanng data required for evaluating conceptual initial mechanical properties tests of beltiine matenal tie-and computer models related to HLW isolation in an envvon- anned from the decommissioned KRB A pressure vessel and

, g, "ated in the UBR test reactor revealed a ment where sgnif, cant heat energy is produced. Coupled-proc-ess modJls are currently limited by the lack of vahdation data major anomaly in relatrve radiaton embrittlement sensitivity.

appropnate for field scales that incorporate relevant trang Poor correspondence of matenal behavior in test vs. power re-processes. Presented in this document is a discusson of provt- actor environments was obsen/ed for the weak test onentation ous nontsothermal expenments. Processes expected to dome (ASTM L C) whereas correspondence was good for the strong nate heat <tnven liquid, vapor, gas, and s >iute flow dunng the experiment are explained, and the conceptual model for noni-sothormal flew and trantgort in unsaturated, fractured rock is g

desenbed Of particular concem is the abihty to conform the hy- "79

' de9rees C in the UBR test reactor. Properties tests before pothestzud conceptual model, specifically, the establishment of UBR irrad4ation revealed a signifcant differerce in 41 J trant+

higher water saturation zones within the host rock around the ton temperature and upper shelf energy level between the ma-heat source, end the establishment of Countercurrent flow con- tenals. However, the matenals exhibited essentially the same dations w' thin the host rock near the heat source. Field experi. radiation embntilemont sensitivity (both onentations), proving mental plans are presented using the Apache Leap Tuff Site to that the anomaty is not due to a basic difference in matenalirra-iilustrate the implementaten of the proposed methodology. Both maton resistances Possible causes of the onginal anomaty and smalbscale prehminary expenments and a long-term experiment tne significance to NRC Regulatory Guide 1.99 are discussed.

NUREGICR-5896: AUXlLIARY FEEDWATER SYSTEM RISK-HUREG/CR-5886: EXPERIMENTAL AND ANALYTICAL INVESTi- BASED INSPECTION GUIDE FOR THE ST. LUCIE UNIT 1 NU-GATION OF THE SHALLOW-FLAW EFFECT IN REACTOR CLEAR POWER GENERATION STATION. PUGH R.;

PRESSURE VESSELS. THEISS,T.J.: SHUM,0.K. Oak Ridge Na- GORE.B.F.; VO.TV Battelle Memorial Institute Pacific North-tional Laboratory. ROLFE,S T. Kansas. Un:v. of, Lawrence, KS- west Laboratory. August 1992. 31pp. 9209240291 PNL 8102.

July 199i. 63pp. 9206060236. ORNL/TM-12115. 62654 001. 63278 237.

The Heavy-Section Steel Technology (HSST) Program is in- In a study sponsored by the U.S. Nuclear Hogulatory Com-vestigating the increase in effectwo fracture toughness of A $33 miss*on (NRC), Pacific Northwest Laboratory has developed and B steei assocated with shallow flaws and the implications of apphed a methodology for deriving plant Specific nsk. based in-the shallow. flaw oficct on reactor pressure vessel (RPV) life as' specton guidance for the auxiliary feedwater (AFW) system at sessments. Test data from beams indicato a significant increase pressurized water reactors that have not undergone probabalistic in the fracture tough'iess of shallow- crack specimens com- nsk assessment (PRA). This methodology uses existing PRA e-pared with deep-crack specimens in the transition region of the sults and plant operating experience information. Existing PRA.

toughness curve for unirradiated A 533 8 steel. If the toughness based inspection guidance inforrr$9n recently developed for increase present in the test specimens were also present to a the NRC for vanous plants was usod to identify genenc compo-reactor vessel, the impact on pressunZed thermal shock (PTS) nont failure modes This info'mation was then comb't.ed with analyses could be signiScant. To facihtate transferability of the plant-specihc and industry-wide component infGrmation and fail-specimen data to an RPV, posttest finite-element analyses have ure data to identify failure modos and failure mechanisms for been performed on several test specimens and a reactor vessel the AFW system at the n?octed plants. St. Lucie Unit 1 was so-for a singte (PTS) transient. The analyses are suthciently rehned lected as one of a senes of pL.nts for study. The product of this to allow interpretaten of the results in terms of the J-integral effori is a pnontzed bsting of AFW imres which have occurred and the so- called O.rtress parameter under plane strain analy- at the plant and at other PWRs. This sting is intended for use sis assumptons. A negatwe O-stress parameter is indicative of by NRC inspectors in the preparatior of inspecton plans ad-a loss of crack tip constraint, which is associated with an in- dressing AFW nsk.important components at tue St Lucie Unit 1 crease in the fracture toughness. Analyses of the test spec 8- plant.

mens indicate that at the onset of crack inriiation the deep-crack specimens exhibit an assentially zero 0- stress parameter NUREG/CR-59n5: REVIEW AND DE/ELOPMENT OF COMMON but that the shallow-crack specimen exhibits a O- stress param. NOMENCLA'.JRE FOR NAMING AND LABELING SCHEMES eter of about -0 7, which indicates a sut)stantial loss of con- FOR FT'C3ABILISTIC RISK ASSESSMENT. TRUSTY,A.Da straint in the shallow-crack beam. Using the test data and post- MACKOWtAK.D.P, EG&G Idaho, Inc. August 1992. 65pp.

test analysis, a locus of toughness data en terms of the J-inte- 9209280052. EGG-2615. 63335 216.

gral and the O-stress parameter has Leen con tructed for a This report desenbes the review and development of commun particular temperatare Ana!yses were also performed on en nomenclature for naming and labeling schemes for probab;listic RPV with a shallow flaw under PTS loading conditions up to the nsk essessments (PRAs) conducted by the Idaho National Engi-maxeim value of J At maximum J, the analyses reveal a O- neenng Laboratory (INEL) Based on the review, the INEL rec-stress parameter about -0.2 to -0 4, which indicates some con- ommends using an exrsting basic event labehng scheme and i

r a 14 - Main Citations and Abstracts e

enisting naming schemes for systems, component types, and ' Pacific North *wst Laboratory conducted a Phase 1 aging as.

component failure modes. The rewow showed no adequate ac- sessment of the standby liqu:d control (SLC) system used in cident sequence labeling schemes currently exis-t. Therefore, . boihng-water reactors. The study was based on detailed reviews the INEL developed a scheme th6i would meet the reiew re- of SLC syste.n component and operatq npenence informae quirements of not exceeding 16 characters and being highly de- tion obtained from the Nuclear Plant Reliability Database senptive of the accident sequence involved. As parts of the de- System, the Nuclear Document System, Licensee Event Re-veloped accident sequence labehng schems, the INEL also de- ports, and other databases. Sources on sodium pentaborate, veloped transient and loss- of-coolant accident initiating event borates. and bonc acid, as well as the effects of environment codes, a sequence naming scheme, and accident type codes, and corrosion in the SLC system were also reviewed to charac- ..

Apphcat ons of the accident sequence labehng scheme are pre- terize chemical properties and corrosion Charactenstics of boral-bented along with tables to allow changes from other schemes ed solutions, Relatively few SLC component failures were attrib-to the recommended naming sAemes. The revow and devel- uted to sodium pentaborate buildup or corrosion. The leading j opment were conducted to peuvide the Nuclear Regulatory aging degradation concern tc date appears to be setpoint dnft Commission w!!h the means to coordinate and integrate their in- in retef valves, which has been discovered during routine sur-ternal octrvities through a common norrenclature for their many veillance and is thought to be caused by mechanical wear. A data bases. higher setpoint resufts in loss of system over. pressure protec-tion, and a decrease in setpoint results in a reduction of boron NUREG/CR 5910: LOSS OF ESSENTIAL SERVICE WATER IN iniection rate. Degradation was also ob,.srved in pump doals LWRS (Gt-153). Scoping Study. CRAMOND,W R.; and internal valves, which could prevent the pumps from oper.

MiTCHELL,0,0 Sandia National Laboratories. YAKLE.JL; et al. ating as required by the technical specG.3tions. In general, Science Appbcations International Corp (formerly Science Ap- however, the results of the Phase I study livhD that age-relata phc9t ons, Inc.): August 1992. 365pp. 920924036L SAND 92- ed degradation of SLC systems has not tun Witous.

1084. 63276 079 The costnbution of essential service water (ESW) system fait NUREG/CR 6003: DENSITY-WAVE INSTABILITIES IN BOILING ure to core damage frequency has long been a concern of the WATER REACTORS. MARCH-LEUBA J. Oak Ridge Naf onal NRC. The objective of this study is to assess the safety signifs Laboratory. September 1992. 54pp. 9210050018. O"INL/TM-cance of the loss of ESW systems in LWRs relative to core 12130. 63382:001.

damage frequency (CDF) and perform a limited value/ impact This report contains a review of issues related to dentity-analysis of potential modificatio a t solve ESW vulnerabilities wave instabihtlos in boihng water reactors (BWRs). The report usseg a prototypica' (pilot) plant. Previous studies indicate that describes the types of instability modes that can be expected in service water systems coMribute from <1% to 65% of the operating reactors. These modes are; (1) the channel thermo-total internal CDF. For the pilot plant analyzed, common ESW hydraube instability mode; (2) the core. wide instabahty mode; vultserabihties are failure of standhy service water pumps to and (3) the out of-phase instabihty modo. The physical mecha-~

stat backflow through check valves for cros* tied pumps, and nisrns leading to each typ6 of instabihty are reviewed and docu failure of normaliy closed isolation va ves in diesel generator mented, along with some ccmmon mathematical models used -

cooling loops to open on osmand. For the potential modifica- in stability Calculations. The main approximations used in these tions evaluated for the pilnt plant, the resufts showed that they mathematical models are presented, and their impact on the ac-could reduce the CDF by as much as 33 percent However, the curacy of the ca!cutations is reviewed. The linear behavtor of a

dollars per person REM measures result ng from vanous groups BWR is studied through the use of transfer functions, and the of these modifications sigmficantly exceeded the current criteria nonhnear behavior and hmit cycle development are studied. A of $1,000. The results, since they only apply to the pilot summary of the sensitivities to physical parameters is also in-plant, are not typical of all BWRs Due to the importance of cluded in this report.

service water to CDF and the plant specific nature of ESW Systems there Could be plants for which there would be cost NUREG/CR 6008: CONSTRAINT EFFECTS ON FRACTU.E .

effective modifications Additional analysis woutd be required TOUGHNESS FOR CIRCUMFERENTIALLY ORIENTED to identify them. CRACKS IN REACTOR PRESSURE VESSELS. BASS.B.R.

SHUM D.K.: KEENEY WALKER Oak Ridge National Laboratory.

NUREG/CR-5930: HIGH INTEGRITY SOFTWARE STANDARDS August 1992. 151p 9209240303. ORNL/TM 12131.

AND GUIDELINES. WALLACEER.; IPPOUTO.LM.; KUHN,D,R, 63272;187& i Natnnal Institute of Standards & Technology (formerly National Pressunzed-thermal-shock (PTS) loading produces biaxial =

Bureau rd Standa Septembc- 1992.10Bpp. 0210050044. NIST stress fields in a reactor pressure vessel (RPV) wall with one of SP 500M 63382:057. the pnncipal Stresses aligned parallel to postulated surface ,

Thia repori presents results of a study of standards, draft cracks in either longitudinal or circumferential welds. The limited standards, and guicetines (all of which wtli hereafter be referred quantity of erlsting beaxial test data suggest a significant de-to as documents) that provwle requirements for the assurance crease of fracture toughness under out of plane (i.e, parallel to .

' of software in safety systems in nuclear power plants. The study the crack front) biaxial loading conditions when compared with -

focused on identifying the attnbutes necestary in a standard for toughness values 'obtained under uniaxial conditions. Any in.

providing reasonable assurance for software in nuclear systems crease in crack tip constraint resulting from these out of-plane The study addressed some issues involvd in demonstrating biaxjal stresses would act in opposition to the in-plane corv conformance to a standard. The documents vary widely in their straint relaxation that has been previously demonstrated for requirements and the precision with which the requirements are shallow cracks. Consequently, understanding of both in-plane ll expressed Recommend.Jons aro provided for guidance ad- and out-of-plane crack tip constraint effects is necessary to a crossirg r.e assurance of high integrity software. It is recom- refined analysis of fracture initiation from shallow cracks under monded that a nuclear hdustry standard be developed based PTS transient loading. This report is the second in a senes in-on the documents reviewed in this study with additiona! atten. vestrgating the potential impact of far field out of plane stresses tion to the concerns identified in *his report. and strains on fracture initiation toughness. Selected fracture prediction models, previously vahdated for small-scale fracture NUREG/CR 4001: AGING ASSOSMENT OF BWR STANDBY specimens under nearly plane strain conditions, were applied to LtOUlO CONTROL SYSTEMS. BUCKLEY,G D.; ORTON.R.D.; additional large scale data with the objective of validating JOHNSONAB ;& 11. Battelle MemoriM Instituts. Pacific North- models ln the plane stress to. plane strain domain before apply-west Laboratory. August 1992 4 top. 9208250215. PNL-8020, ing them to positive out of plane strain conditions. The general 63037.217. finding was that apphcations of the models resulted in predio-

, . _..--_-....u.--_....________a.__.____.-...-_._.a_,.--

_. - .- - - . - - . - - , . - ~ . .- . . ~. .. .

Main Citations and Abstracts 15

' bons of fracture behavior that confiscled with existing experb supplemental information on the thormal sleeve probiemcwhich mental data considered relevant to the problem. Because of the affected Westinghouse nuclear plants using Generation 3 ther.

confheting results, it is apparent that testing of RPV steels is re- mal sleeven. The informabon presented is intended to be the Quired (1) to determine the magnitude of out of. plane biasal basis for deciding whether the issue can be considered re-

- foading effects on fracture toughness; and (2) to provde a basis solved or if addtbonal information is required to resolve it.

for development of predictive models. This course of act on is r ocessary to support a refined treatment of in plane and out of- NUREG/lA-0042: DISPERSED FLOW FILM BOILING.An invesu-plane constraint effects in PTS analysis Proposed in this report gat on Of The Possibihty To improve The Models implemented are cntena for a biaxial specimen that would forrn the basis of a in The NRC Computer Codes For The Reflooding Phase Of The LOCA, ANDREANI M. Paul Scherror insbtute. ANDREANI,M

• teshng program designed to provide data to explain dif,arences YADIGAROGLU,G.; et al Swiss Federal institute of Technologh t etween theoretical predictions and measured matenal behav-ior. Results of design studies on the biaxiat specimen wdl be (ETH). August 1992, 68pp. 9209220485. 63261:206 presented in a future report from the Heavy.Soction Steel Tech- Dispersed Flow Flim Boang es the heat transfer regime that nology Program. occurs at high void fract ons in a heated channet The way this heat transfer mode is modelled in the NRC computer codes NUREG/CR-6009 V01: DEVELOPING AND ASSESSING ACCI- (RELAPS and TRAC) and the valdity of the assumptions and -

DENT MANAGEMENT PLANS FOR NUCLEAR POWER empinct.: correlat ons used is discussed. An extensive review of -

PLANTS. Development Process ' And Criteria. HANSON.D.J.; the theoretical and expenmental work related with heat transfer BLACKMAN,H S.; MEYER,0.R ; et at EG&G idaho, Inc. August to highly dispersed mixtures reveals the basic deficiencies of 1992. 78pp. 9209240241. EGG-2682. 63273.254- these modelt the investigation refers mostty to the typical cord This document is the first volume of a two volume report. It dttions of low rate bottom reflooding, since the simulabon of this describes a four phase approach for developing cntena that caG physical situaton by the computnr codes has often showed be used for assessing the adequacy of severe accdont man- poor results. The attomative models that are available in the fit.

agement plans for nuclead power plants. The general attnbutes erature are reviewed, and their monts and limits are highlighted.-

of accident managernent plans (Phase 1) are identfied, and a The modificabons that could improve the physics of the models process for developing and implementng severe accident man- implemented in the codes are identihed.

agement plans (Phase 2) is desenbed. This process is based on a prototype process described in NUREG/CR-5543. The proto, NUREG/lA-0054: ASSESSMENT OF RELAPS/ MOD 2. CYCLE type process was revised using results from an evaluation-of 36.02, USING NEPTUN REFLOODING EXPERIMENTAL DATA.

this process (Phase 3), which is documented in Volume 2. Gerp RICHNER Ma ANALYTIS,0.Ta AKSAN,S.N. Paul Scherrer Insti.

. eral critena for assessing the adequacy of accdont manage. tute. August 1992,105pp. 9209240217,63271:211.

ment plans are also presented (Phase 4). These enterta were Seven NEPTUN reflooding expenments with varying param-based on process specific cntena presented in Volume 2 and eters flooding rate, single rod power, pressure and initial rod NUREG/CR-5543. temperatures were simulated with the code RELAPS/ MOD 2..

vorsion 36.02, to assess the code, especially its reflood model NUREG/CR-6009 V02: DEVELOPING AND ASSESSING ACCl- These calculabons were performod with the specific objectves DENT MANAGEMENT PLANS FOR NUCLEAR POWER of evaluating the general predicbon capability as weil as specific PLANTS Evaluation Of A Prototype Process. HANSON,D.J4 problem areas of the RELAPS/ MOD 2 code in modelling boil off JOHNSON,S.Pa BLACKMAN H.S.; et c'. EG&G Idaho, Inc. July and reflood behavior, The differences between code predictions 1992.182pp. 9207270263, EGG-2682. 62509:178. and expenments are desenbed and analy2ed. Implement ng new This document is the second of a two-volume report that dis- correlations into the code and modifving of correcting existing cusses development of accident management plans for nuclear correlations, for exarnple for wall heat transfor or interphase power plants. The first volume: (a) desenbes a four-ph ise ap- friction, some of the weak points of the code during reflooding proach for developing criteria that could be used for assessing could be identified.

the adequacy of accident management plans; (b) Identifies the General attnbutes of acc dent management plans (Phase 1); (c) NUREG/tA-0067: RECIRCULATION SUCTION LARGE BREAK .

presents a prototype process for developing and implementing LOCA ANALYSIS OF THE SANTA MARIA DE GARONA NU-severe accident management plans (Phase 2); and (d) presents CLEAR POWER PLANT USING TRAC-BF1(G1J1). LOPEZ,J.V.

critena that can be used to assess the adequacy of accident Polytechnic Univ. of Madnd, Madrid, Spairt CRESPO J.L.Canta-management plans. This volume: (a) describes results from an bna. Univ. of. Spain. FERNANDEZAA. Nucienc', S.A. (Spain) evaluation of the capabilities of the prototype process to August 1992.63pp.9208250167 ICSP.GA-LOCA T. 62872:223.

produce an accident management plan (Phase 3), and (b), A best estmate analysis of a recirculation suction pipe large based on these results and preliminary enteria inciuded in break loss-of coolant accident analysis for Santa Maria De NUREG/CR 5543, presents modifications to the entena where Garona nuclear power plant using TRAC-BF1 code is present-appropnate. ed. '

NUREG/CR-6010: HISTORY AND CURRENT ST/iTUS OF GEN- NUREG/lA-0088: POST-TEST ANALYSIS AND NODAll2ATION ERATION 3 THERMAL SLEEVES IN WESTINGHOUSE NUCLE. STUDIES OF OECD LOFT EXPERIMENT LP 02-6 WITH

l. AR POWEH PLANTS. MARTIN.G. MARTIN Consulting Services, RELAPS/ MOD 2 CY36 02. LUBBESMEYER,0. Pauf Scherrer in-l- Inc.

• S Cohen & Associates, Inc. July 1992.' 69pp. stitute. August 1992.186pp. D209240233 PSI-BERICHTNR92.

9208250214. 62871:00t. 63272:001, I

- From rnid-1982 until 1987, loose tnermal sleeves or sleeves This report presents the resutts and analysis of nine post-test with cracked attachment welds were found in several opersbng calculations of the Experiment LP 02 6 by using RELAPS/

Westinghouse nuclear power plants. Westinghouse investiga. MOD 2 CY36-02 computer code with different nodalizations..

tions concluded that these occurrences had been confined to Starting with a ." standard nodalization" we have reduced the . ~

those thermal sleeves of " Generation 3" design. The sleeve number of volumes and junctions as well as the number of ,

problem was a genenc issue (Generic issue No. 73) but affect- radial zones in the fuel rods, for dfferent nodalization studies.

ed only those plants using Generabon 3 sleeves. The NRC's Except for the claddng temperatures, only small discrepancios -

Safety Evaluation Report, " Evaluation of Thermal Sleeve Prob- have been observed for the other main parameters of the re-lems in Westinghouse Plants", dated October 28,1983c con- suits of runs using different nodalizations but reduced number tains a proposed resolution for the thermal sleeve issue. Be- of volumes and junctions usually have lead to a decreased run--

cause a resolution had been proposed, the issue was consid- ning time for the problem. The time behavtors of the cladding ered to be nearly resolved. This report presents updated and - temperatures have been significantly affected by the chosen no-l-

p

- _ _ _ _ . _ . ~ _ _ _ _ _ _ _ _ , .

7 16 Main Citations and Abstracts '

daluabons, The most comparable results with the egenmental the other LOFT large break expenments, RELAP5/ MOO 2 was data t,we been schwed by using medium number of nodes not able to predict this phenomenon except with a Certain ma-With respect to high mass flux, carty bottom-up rewetting, one nipulation by initiatog the reflood option.

of the key-events of Expenment LP-02 6 as well as of most of _

Secondary Report Number index This index lists, in alphabetical order, the performing organization-issued report codes for the NRC contractor and international agreement reports in this compilation. Each code is cross--

referenced to the NUREG number for the report and to the 10-digit NRC Document Control System accession number.

l l

1 REPORT NUMBER SECONDARY REPORT NUMBER REPORT NUMDER SECONDARY REPORT NUMBER ORNL/ T M-11049 NURE GIC A-5872 l ANL 91/22 NUREG/CR4744 V06 N1 ORNL/TM-11606 NUREG/CR 5700 ANL 92/3 NUREG/CH 5840 DANL/1M 11969 NURE G /CR-5793 ANL 92/30 NUREGICR-4667 V14 O R /

BHARC700/91/014 NUREG/CR-5758 V02 / 02 BMI-2173 NUREGICR.4599 V02 N1 ORNL/TM 12073 NURE'G/CR-5859 DNL NUREG 51581 NUREG/CR.2907 VIO ORNL/TM-12115 NUR( G/CR-5886 BNL NUREG 51934 NUREG/GR 4409 VO4 ORNL /TM 12130 NUREG/CR-6003 NUREG/CR 5819 ORNL/I M- 12131 NUREG/CR f008 BNL NUREG-52313 NORE G/CR-5867 CNWRA 91014 NUREG/CR 5804 ORNLSUG797778'O PNL-5210 NUREG/CR-3950 V07 EGG-2567 NUREG/CR4378 NUREG/CR 4469 V13 NURE G/CR-5416 PNL-5711 EGG-2571 PNL-5711 NUREG/CR-4469 V14 EGG-2015 NUREG/CR-5905 PNL 7/36 NUREG/C505758 V02 E GG 2626 NUREG/CR4356 V02 P /,

E GG-2626 NUREG/CR4356 V01 E GG-2655 NUREG/CR 5646 PS3 BERIGHTNR92 NUREG/lA-0086 EGG 2657 NUREG/CH 5767 V01 SAIC-92/1137 NUREG/CR 5587 EGG ?ti57 NUREG/CR-5787 VU2 SAND 07 7156 NUREG/CH-4839 EGG-2676 NUREG/CP-0123 SANO90-0087 NUREG/CR 54 3 NUREG/CR4391 SAND 901022 NUREG/CR-8564 EGG 2600 NUREG/CR4305 V01 EGG-2682 NUREG/CR4009 VO2 SAND 90-2765 SAND 91 1535 NUREG/CR 5790 E GG-2682 NUREG/CR 6000 V01 NUREGICR-5772 V01 NUREG-1442 RO1 SAND 91 1766/1 FE MA REP-17 SANO92 0173 NURL O/C$t0120 ICSP GA-LOCA T NUREG/1A 0067 SAND 92-0537 NUREG/CR-4832 VC1 LA-12031 M NUREG/CR-5673 V03 53 LA 12031-M NUREG/CR-5673 V02 h 337 U 48 MEA 2458 NUREG/CR-5891 SAND 924537 NUREG/CR-4632 VO3 P1 NIST SP 500 204 NUREG/CR-5930 SANO92 0537 NUREG/CR-4832 V03 P2 ORAU-92/C57 NUREG/CR 5849 ORF FC SANO921084 NUREG/CR-5910 ORNL 6687 NUREG/CR-5779 Voi sRA.9201 NUREG/CR 5a68 NUREG/CR-4674 VIS UCRL40109798 NUREG/CR 5864 ORNL/140AC-232 NUREG/CR-3320 V02 ORNL/NOAC-232 NUREG/CR4674 V16 WHCrEP4204 17

. . _ .__ ., _ .._.._ ~ -- - - , ._ __ _.__

2..t-.+M4 e '

- e4h 4 --__.4y a h44._m 3 +w-4a. AmA a 4A_,--m-mp_w_. ge- s a m A3_pa._g.--m ,y%m.a% . gu,_m.am,,,, _g._,,m._, , _y_,,, _ ,, ,, ,4 ,. _3_,,, , ,

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.......--...-.--...-.____.-,__,-~..L.____....,.__._...._..___.__-___m.___ _ _ _ . , _._-__ . . . _ _ - _ _ .- _.___ _.-_.. -- . , , -_ , .m.,,,

Personal Author index This index lists the personal authors of NRC staff, contractor, and international agreement reports in alphabetical order. Each name is followed by the NUREG number and the title of the report (s) prepared by the author, if further information is needed, refer to the main cita- l tion by the NUREG number.

\

BERGER,JL AKSAN.S.N.

NUREG/lA.0054- ASSESSMENT OF RELAP5/ MOD 2. CYCLE 36 02. NUREG/CR5849 DAF FC MANUAL FOR CONDUCTING RADIOLOGi-USING NEPTUN REFLOODING EXPERIMENT AL DATA CAL SURVEYS IN SUPPORT OF LICENSE TERMfNATION Draft ALLENSPACH.F.

NUREG 1214 R10 HISTORICAL DATA SUMM ARY OF THE SYSTEMAT- BERTING,F.M.

IC ASSESSMENT OF LICENSEE PERFORMANCE. NUREG/CR 3950 V07: FUEL PERFORMANCE ANNUAL REPORT FOR 1989-ANALYTIS,0.T.

NUREG/tA 0054 ASSESSMENT OF RELAP5/ MOO 2, CYCLE 36 02. BLACKFORD,ML USING NEPTUN REFLOODtNG EXPERIMENTAL DAT A.

NUREG 1451: STAFF TECHNICAL POSITION ON INVESTIGATIONS TO i ANDREANI,M- IDENTIFY FAULT DISPLACEMENT HA2ARDS AND SEISMIC HA2 NUREG/lA 0042 OtSPERSED FLOW FILM BOiltNG.An hevostigaten 01 ARDS AT A GEOLOGIC REPOslTORY.

The Posstakty To impre te The Models implemented in The NRC Com-puter Codes for The Refloodng Dbase Of The LOCA. BLACKMAN.H.S.

NUREG/CR-6009 V01. DEVELOPING AND ASSESSING ACCOENT MANAGEMENT PLANS FOR NUCLEAR POWER NUR R-Se54. UNIVERSAL TREATMENT OF PLUMES AND D' D 8 STRESSES FOR PRESSURIZED THERMAL SHOCK EVALUATIONS. A D ASSESSING ACCIDENT MANAGEMENT PLANS FOR NUCLEAR POWER PLANTSEvaluaton ATWOOD,C.L Of A Prototyoe Process.

NUREG/CR-5378. AGING DATA ANALYSIS AND RISK ASSESSMENT.

DEVELOPMENT AND DEMONSTRATION STUDY. BLOSE.R.E.

BAILEY,W.J. NUREG/CR5443. CORE CONCRETE INTERACTIONS USING MOLTEN NUREG/CR-3950 V07. FUEL PERFORMANCE ANNUAL REPORT FOR URANIA W1TH ZlRCONIUM ON A LIMESTONE CONCRETE WD BASEMAT.The SURC 1 Expenment NUREGICR 5564 CORE-CONCRETE INTERACTIONS USING MQLTEN

^ ^ ^ ^

NU E'G/CR-4832 V07: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR E"E*"* "

POWER PLANT: RfSK METHOOS INTEGRATION AND EVALUATION PROGRAM (RMIEP) External Event Scoomo Quannfeatori BORKOWSKt'J A' NUREG/CR4839. METHODS FOR EXTEMNAL EVENT SCREENING NUREG/CR4356 V01: TRAC-BF1/ MOD 1: AN ADVANCED DEST-ESTp OUANTIFICATION- RISK METHOOS INTEGRATION AND EVALUA. MATE COMPUTER FROGRAM FOR BWR ACCIDENT TION PROGRAM (RMIEP) METHODS DEVELOPMENT, AN ALYSIS Model Description.

BARTHOLD,W.P. NUREG/CR-4391: TRAC /BF1. MOD 1 MODELS AND CORRELATIONS.

NUREG/CR5010 EVALUATION OF MHTGR FUEL REllADILITY.

BROCKMANN,JL B ASS,B.R- NUREG/CR-5443: CORE CONCRETE INTERACTIONS USING MOLTEN NUREG/C45793- A COMPARISON OF ANALYSIS METHODOLOGIES URANIA - WITH ZlRCONIUM ON A LIMESTONE CONCRETE BASEMAl.The SURC-1 Expenment FOR ACTORPREDtCTING PRESSURE VESSEL CLEAVAGE SUBJECTED TOARREST OF A DEEP CRACK PRESSURIZED-THER- NUREG/CR IN A RE'CORE-CONCRETE 5564: . INTERACTIONS USING MOLTEN

^ ^

NU b 72 NZ A Fi ELEMENT MESH GENERATOR "" #

FOR NOZ2LE-CYLINDER INTERSECTIONS CONTA!NING INNER-CORNER CRACKO BROWM NUREG!CR6000: CONSTRAINT EFFECTS ON FRACTURE TOUGH- NUREG/CR5305 VOI: INTEGRATED RISK ASSESSMENT FOR LA-NESS FOR CIRCUMFERENTIALLY ORIENTED CRACKS IN REAC SALLE UNIT 2 NUCLEAR POWER PLANT Phenomenology And Risk TOR PRESSURE VESSELS Uncertainty Evaluation Program (PRUEP).

BATES.G.

NUREG 1442 RO1: EMERGENCY RESPONSE RESOURCES GUIDE For BRUST.F.

Nuclear Power Plant Emergencies. NUREG/CR4599 V02 N1. SHORT CRACKS IN PIPING AND PIPtNG WELDS. Semiannual Report. Aprs-September 1991..-

DAUM,J.W.

NUMEG/CR4409 V04: DATA BASE ON DOSE REDUCTION RE- BUCKLEY,0.D.

SEARCH PRO.lECTS FOR NUCLEAR POWER PLANTS- NUREG/CR 6001: AGtNG ASSESSMENT OF BWR STANOCY LtOUID CONTROL SYSTEMS.

BENNETT,T.E.

NUREG/CR 5068 DEVELOPMENT OF POSITION SENSITIVE PROPOR.

T NA NT R F HOT PARTICLE DETECTION IN LAUNDRY BURSON.S.B[55 NUREG 14 DRFT FC: ACCIDENT SOURCE TERW F

~

WATER NUCLEAR POWER PLANTS. Draft Report For Comment.

BERANEK.A.F. CHANIN,D.L NUREG/CP 0122 V01: PROCEEDINGS OF THE AGING RESEARCH IN_ NUREG/CR5305 V01; INTEGRATED RISK ASSESSMENT FOR LA-FO 1MATION CONFERENCE SALLE UNIT 2 NUCLEAR POWER PLANT,Pnenon'enology And Risk NUREG/CP-0122 V02 PROCEEDINGS OF THE AGING RESEARCH IN-FORMATION CONFERENCE. Oncertamty Eva;uabon Program (PRUEP) 19

_ _ . . . . - __ .. . ~_ _ _ _

$0 Personal Author Index CHOPRA.O X DINGMANA NUREG/CR4744 %06 N1: LONG TERM EMBRITTLEMENT OF CAST DUPL,EX STAINLESS STEELS tN LWR SYSTE MS.Sermannual NUREG/CP-0124 WORKSHOP ON THE USE OF PRA METHODOLOGY ReporLOctober 1990 March 1991. FOR 1HE ANALYSIS OF REACTOR EVENTS AND OPERATONAL DATA.

CHOWDHURY,A.H-OINGMAN S.E.

NUREG/CR 5804 REFOSITORY OPERATIONAL CRITERIA ANALYSIS NUREG/CR4832 V03 Pt ANALYSIS OF THE LASALLE UNIT 2 NU-CHUNG.H.M. CLEAR POWER PLANT: RISK METHOOS INTEGRATON AND EVAL-NUREG/CR-4667 V14. ENVIRONMENTALLY ASSISTED CRACKING IN UATION PROGRAM (RMIEP)lnternal Events Acenient Sequence LIGHT WATER REACTORS Serrmannual Report. October 1991 March QuantAcaten Mam Report.

1992. NUREG/CR4832 V03 P2; ANALYSIS OF THE LASALLE UNIT 2 NU-CLETCHE R.J.W. CLEAR POWER PLANT: RISK METHOOS INTEGRATION AND EVAL-UATION PROGRAM (RMIEP) interna! Events Accdent Sequence NUREG/GR 46?4 V15 PRECURSORS TO POTENTIAL SEVERE CORE Quant:ficatutAppereces DAMAGE ACCIDENTS 1991 A STATUS REPORT Men Report And Appendm A.

DOCTOR.SA NUREG/CR4674 V16 FRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1991 A STATUS REPORT.Apper@ces B.C. NUREG/CR4469 V13' NONDESTRUCTIVE EXAM;NATON (NDE) REll.

And & ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER REACTORS.Somiannual Report. October 1990 March 1991.

CONGEMI,J. NUREG'CR4469 Vi4: NONDESTRUCTIVE EXAMINATION (NDE) REll-NUREG/CR 2907 V10 RADtOACTIVE MATERIALS RELEASED FROM ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER NUCLEAR POWER PLANTS Arnual Report 1989 REACTORS Semsannual Report Apnl 1991. September 1991.

COPINGER,0.A. DOLAN.B.W.

NUREG/CR4674 VIS: PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4674 V15: PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCOENTS- 1991 A STATU3 REPORT Mam Report And DAMAGE ACCIDENTS: 1991 A STATUS REPORT Mam Report And -

Appends A. Apperds A, NUREG/CR 4674 V16 PRECURSORS TO POTENTIAL SEVERE CORE NUREG/CR4674 V16. PRECURSORS TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1991 A STATUS REPORT.Appereces B.C. DAMAGE ACCOENTS 1991 A STATUS REPORT. Appendices B.C.

And D. And D.

COPUS.EA DURBIN,N.

NUREG/CR 5443 CORE CONCRETE INTERACTONS USING MOLTEN NUREG/CR 5758 V02: FITNESS FOR DUTY IN THE NUCLEAR POWER URANIA WITH ZlRCONIUM ON A LIMESTONE CC 4 CRETE INDUSTRY. Annual Summary Of Program Performance ReporttCY BASEMAT.The SURC-1 Expenment i gg , '

NUREG/CR 5564: CORE-CONCRETE INTERACTONS USING MOLTEN UO(2) WITH ZlRCONIUM ON A BASALTIC BASEMAT TM SURC 2 E xpenment- EMRITA NUREG-0933 $14; A PR!ORIT12ATION OF GENERIC SAFETY ISSUES.

CHAMOND,WA NUREG/CR-5910- LOSS OF ESSENTIAL SERVICE W ATER IN LWRS EVANS,0.D.

(GM53) Scoping Study NUREG/CR 5880- NONISOTHERMAL HYDROLOGIC TRANSPORT Ex-CRESPO,JL PERIMENTAL PLAN NUREG/lA4067 RECIRCULATON SUCTON LARGE BREAK LOCA FADDEN.M.

ANALYSIS OF THE SANTA MARJA DE GARONA NUCLEAR POvIER NUREG-0525 V00 SAFEGUARDS

SUMMARY

EVENT LIST (SSEL). Pre.

PLANT USJNG TRAC-BF1(G1J1). NRC Through December 31,1980 NUREG 0526 V02: SAFEGUARDS

SUMMARY

EVENT LIST NURE R 5685 SEALING PERFORMANCE OF BENTONITE AND BENTONITE / CRUSHED ROCis BOREHOLE PLUGS FERNANDEZ,RA NUREG/CR-568 7-. BOREHOLE STABILITY IN DENSELY WELDED TUFFS' NUREG/tA4067; RECIRCULATION SUCTON LARGE BREAK LOCA DANIEL,S, ANALYSIS OF THE SANTA MARIA DE GARONA NUCLEAR POWER PLANT USING TRAC-BF1(G1J1)

NUREG/CR.5790t R!SK EVALUATON FOR A B&W PRESSUR12ED WATER REACTOR, EFFECTS OF FIRE PROTECTON SYSTEM AC. FERRELL,C.M.

TUATON ON SAFETY-RELATED EQUIPMENT.Evatuation Of Genene NUREG-1465 DRFT FC: ACCOENT SOURCE TERMS FOR LtGHT.

Issue 57 WATER NUCLEAR POWER PLANTS. Draft Report For Comment DANIEL,SL .

FIELD,L NUREG/CR4832 V03 Pt ANALYSIS OF THE LASAI LE UNIT 2 NU- NUREG/CR 5758 V02: FITNESS FOR DUTY IN THE NUCLEAR POWER CLEAR POWER PLANT: RISK METHOOS INTEGRATION AND EVAL- INDUSTRYAnnual Summary Of Program Performance Reports.CY UAT10N PROGRAM (RM EP).Intemal Events Accident Sequence 1991 Quanblication.Mam Report 4

NUREG/CR-4832 V03 F? ANALYSIS OF THE LASALLE UNIT 2 NU- FLEMING,T.

CLEAR POWER PLANT: RISK METHOOS INTEGRATON AND EVAL-UATION PROGRAM (RMIEP)lnternal Events Accdont Sequence NUREG/CR-5758 V02: FITNESS FOR DUTY IN THE NUCLEAR POWER Ouantfication.Apper@ces. INDUSTRY.Ar.nuat Summary Of Program Performance Reports.CY '

1991.

DEDHIA,D.D.

FRANCINtA NUREG/CR-5064. THEORETICAL ANO USER'S MANUAL FOR PC-

. PRAISE.A Probabilisuc Fracture Mecharucs Computer Code For Piomg NUREG/CR4599 V02 N1: SHORT CRACKS IN PIPING AND PIPING Rehabihty Analysis. WELDS Serniannual Report. Aprii-September 1991.

Ot&MOND.D.J. FR1 LEY,JA NUREG/CR 5819. PROBABILITY ANO CONSEQUENCES OF RAPID NUREG/CR4469 Vid NONDESTRUCTIVE EXAMINATION (NDE) REU-BORON DtLUTION IN A PWR.A Scopeg Study ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER DIQZ,A A REACTORS SeMannual Report Aoni 1991-Septernber 1991, t.

NUREG/CR-4469 Vt4 NONDESTRUCTIVE EXAMINATION (NDE) RELI. FUENKAJORN,K.

ABILfTY FOR- INSERVICE INSPECTON OF LIGHT WATER REACTORS Sermannual Report Apnl 1991. September 1991. NUREG/CH-5687: BOREHOLE STA88UTY IN UENSELY WELDED TUFFS.

m _ _ _ _

p; Personal Author index. 21 GEHL.A.C, NUREG/Cre4469 V14: NONDESTRUCTIVE EXAMINATON (NDE) REU-NUREG/CR-5700; AG!NG ASSESSMENT OF NEACTOR INSTRUMEN- ADtUTY FOR INSERVICE INSPECTION Of UGHT WATER T ATON AND PROTECTON SYSTEM COMPONEN1S.Agmg Rc'sted REACTORS.Semannual Report April 1991. September 1991.

Operatog Experences.

HUGHEY.CL GHADIAU.N. NUREG/CP-0120 PROCEEDtNGS OF THE FIFTH WORKSHOP ON -

NUREG/CR4599 V02 N1- SHORT CRACKS IN PIPING AND PIPING CONT AINMENT INTEGRITY. Held in Washington,0C May 12-14,1992.

WELDS Semannual Report, Apni September 1901.

IBRAHIM,A.B.

GILES,M.M- NUREG-145t; STAFF TECHNICAL POSITON ON IfuE0TIGATONS TO NUREG/CR-4356 V01 TRAC oft / MOD 1: AN ADVANCED BEST ESTb IDENTIFY FAULT DISPLACEMENT HAZARDS AND SEISMIC HAZ-MATE COMPUTER PROGRAM FOR BWR ACCIDENT ARDS AT A GEOLOGIC REPOSITORY.

ANALYSIS Model Desenption.

NUREG/CR4356 V02: TRAC-OF1/ MODI:AN ADVANCED BEST ESTI- iPPOUTOAM.

MATE COMFU1ER PROGRAM FOR BOILING WATL ' REACTOR AC- NUREG/CR5930 HIGH INTEGRITY SOFTWARE STANDARDS AND CIDENT ANAD SIS User's Guide GUIDEUNES.

GOLD R.

NUNEG/CR3320 V02. LWR PRESSURE VESSEL SURVEltLANCE 00-S; METRY lMPROVEMENT PROGRAM PSF S.artup Expenments R'EG CRE867. GRADIENT STUDY OF A LARGE WELD JOINING TWO FORGED A 506 SHELLS OF THE MIDLAND REACTOR GOOD,N.S. VESSEL NUREG/CR4469 V11 NONDESTRUCTIVE EXAMINATON (NDE) REU-JACOBUS,M.J.

ABIU TY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS Semiannual Report October 1990 March 1991. NUREG/C45772 V01- AGING. COND' TION MONtTORING. AND LOSS-NUREG/GR-4469 V14 NONDESTRUCTIVE EXAMINATON (NDE) REU. OF COOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E ELECTRICAL AuluTY- FOR INSERv1CE INSPECTON OF UGHT WATER CABLES Crossknked Potyolefm Cat'es.

REACTORS Semannual Report Apel 1991-September 1991.

GORE,B F. . NUREG/CR4356 V02- TRAC DF1/ MOOT AN ADVANCED BEST-ESTi-NUREG/CA 5996. AUXILIARY FEEDWATEH SYSTEM RISK BASED IN- MATE COMPUTER PROGRAM FOR BOILING WATER REACTOR AC.

SPECTON GUlOE FOR THE ST. LUCIE UNIT 1 NUCLEAR POWER CIDENT ANALYSIS. User 3 Guide GENERATON STATON JOHNSON.A.8, GREENWOOO.M.S. NUREG/C46001: AGING ASSESSMENT OF BWR STAND 0Y LIQUlO NUREG/CP-4469 V14 NONDESTRUCTIVE EXAM: NATION (NDE) REU' CONTROL SYSTEMS.

ABIUTY FOR INSERVICE INSPECTION OF UGHT WATER REACTORS.Semannual ReportApru 1991. September 1991. JOHNSON.J D.

NUREGICR5305 V01: INTEGRATED RISK ASSESSMENT FOR LA-

~

NURE R 5673 V03 TdAC-PF1/ MOO 2 CODE MANUALPro0ramtner's Uncerta nty Evaluahon Program (PRUEP).

.g JOHNSON.S.P.

HAGEMAN.J.P NUREG/CR$804 REPOSITORY OPERATONAL CRITERIA ANALYSIS. NUREG/C46009 V02. DEVELOPING AND ASSESSING ACCOENT MANAGEMENT PLANS FOR NUCLEAR POWER PLANTS Evaluation HAGEN,E.W. Of A Prototype Process.

NUREG/CR-5700 AGING ASSESSMENT OF REACTOR INSTRUMEN.

TATON AND PROTECTON SYSTEM COMPONENTS Aging-Related JONES.K.R-Operaung Expenences. NUREG/CR-5787 V01; itMING ANALYSIS OF PWR FUEL PIN FAILURESFnal ReportMain Text And Appendices Ad HANSON.D.J. NUREG/CR.5757 V02: TIMING ANALYSIS OF PWR FUEL PIN NUREG/CR6009 Vol. DEVELOPING AND ASSESSING ACCIDENT FAlLURES Final Report Appendices K.L MANAGEMENT PLANS FOR NUCLEAR POWER Pt AN'S. Development Process And Cnterg KASSNER,T.F.

NUREG/CR6009 V02: DEVELOPING AND ASSESSING ACCIDENT NUREG/CR4607 V14: ENVIRONMENTALLY ASSISTED CRACKING IN MANAGEMENT PLANS FOR NUCLEAR POWER PLANTS Evaluaten UGHT WATER REACTORS. Semaneual Report. October 1991 - March Of A Prototype Process- 1992.

HARRIS D.O. KATSMA K.R.

NUREG/CR5864: THEORETICAL AND USER'S MANUAL FOR PC- NUREG/C45707 Vol: TIMING ANALYSIS OF PWR FUEL PtN PRAISE.A Probabstistic Fracture Mechamc5 Computer Code For Ppn9 F AILURES Final Report Man Text And Appendees A4 Rehab*ty AnalyM - NUREG/CR5787 V02: TIMING ANALYSIS OF PWR FUEL P1N FAILURES Final Report Appendices K.L HAWTHORNE,J.R.

NUREG/CR 5891; ACCELERATED IRRADIATON TEST OF GUNDREM-KEE'tEY WALKER MINGEN RE ACTOR VESSEL TREPAN MATERIAL NUREG/CR 5793. A COMPARISON OF ANALYSIS METHODOLOGIES HEASLER.P.C FOR PREDICTING CLEAVAGE ARREST OF A DEEP CRACK IN A RE-ACTOR PRESSURE VESSEL SUBJECTED TO PRESSURIZED-THER-NUREGICR4469 V13: NONDESTRUCTIVE EXAMINATION (NDE) REU.

ABluTY FOR INSERVICE INSPECTON OF LIGHT WATER MAL SHOCK LOADING CONDITIONS.

REACTORS Sem6 annual Report October 1994Maech 1991 NUREG/CR-S872: ORNOZL: A FINITE-ELEMENT MESH GENERATOR NUREG/CR-4469 V14. NONDESTRUCTIVE EXAMINATON (NDE) REU. FOR NOZZLE.CYUNDER INTERSECTONS CONTAINING INNER ABluTY FOR INSERVICE INSPECTION OF LIGHT WA TER CORNER CRACKS RE ACTORS Semannual Report Apn1 1991-September 1991; NUREGICM006. CONSTRAINT EFFECTS ON FRACTURE TOUGH-NESS FOR CIRCUMFERENTIALLY ORIENTED CRACKS IN REAC-HIGGINS.S.J. TOR PRESSURE VESSEL &

NUREG/C45305 V)l: INTEGRATED RISK ASSESSMENT FOR LA.

SALLE UNIT 2 NUCLEAR POWER PLANT.Pnenomenology And hsk K H A N,Y.A.

Uncertanty Evaluation Program (PRUEP) NUREG/CR4409 V04. DATA BASE ON DOSE REDUCTION R E-SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

HOCKEYAL

. NUREG/CR4469 V13. NONDESTRUCT!VE EXAMINATON (NDE) REU- KIUNSKI,T.

ABluTY FOR INSERVICE INSPECTION OF UGHT WATER NUREG/CR 4509 V02 N1 SHORT CRACKS IN PIPING AND PtPING REACTORS Semannual Report October 1990 March 1991 WEL DS Serrnannual Report, April-September 1991.

...., - ,_ . . . _ - - , _ - ~ ,, - , - , , - . . . . . _ _ . - . ~ . .- _ . . . . - . _

i 12 -Personal Author Index

- KLAMERUS E. . MACdOWIAK,D.P.

1 NUREG/CR.5790- RISK EVALUATION FOR A B&W PRESSURIZED NUREG/CR-5905: REVIEW AND DEVELOPMENT OF COMMON NO -

' WATER REACTOR. EFFECTS OF FIRE PROTECTION SYSTEM AC-' MENCLATURE FOR NAMING AND LADELING SCHEMES FOR TUATON ON SAFETY-RELATED EQUIPMENT. Evaluation Of Genenc PROBABILISTOblSK ASSESSMENT.

Issue 57.

MAJUMDAR S.

REG'/CR-5819' PROBABILITY AND CONSEQUENCES OF RAPID BORON DILUTION IN A PWR.A Scrpng Study-

^ '""*"""#

• K ONDtC,N.N.

NUREG-1*77 R03. NRC RESEARCH PROGRAM ON PLANT AGfNG- M ARCH-LE UBA).

LISTING AND SUMMARIES OF REPORTS ISSUED THROUGH JULY NUREG/CR 6003. DENSITY-WAVE INSTABILITIES IN BOfLING WATER 1992. REACTORS.

KRtSHNASW AM Y,P. MARSCHALL,C.W.

NUREG/CR-4599 V02 N1; SHORT CRACKS IN PIPING AND PIPING NUREG/CR4599 V02 N1 SHORT CRACKS IN PlPING AND PIPtNG WELDS Semsannual Report, Apr2. September 1991. WELDS. Semiannual Report, Apni-September 1991.

KR YTE R.R.C. MARTIN,G.

NUREG/CR4319 V02- AGING AND SERVICE WEAR OF SOLENOtD' NUREG/CR 601e HISTORY AND CURRENT STATUS OF GENERA-OPERATED VALVES USED IN SAFETY SYSTEMS OF NUCLEAR TION 3 THERMAL SLEEVES IN WESTINGHOUSE NUCLEAR POWER POWER PLANTS Evaluation Of Morutonng Methods.

PLANTS KUHN.DA NUR G/CR 5 0: HIGH INTEGRITY SOFTWARE STANDARDS AND NU E 5673 VC2- TRAC-PF1/ MOD 2 CODE MaNUALUser's Guide.

KULLBERG,C.L MCCONNELL K.L NUREG/CR4331: TRAC /BF 1. MOD 1 MODELS AND CORRELATONS NUREG-1451: STAFF TECHNtCAL POSITION ON INVESTIGATIONS TO IDENTIFY FAULT DISPLACEMENT HAZARDS AND SEISMIC HAZ-KURTZ.R.J-ARDS AT A GEOLOGO REPOSITORY.

NUREG/CR4469 V14 NONDESTRUCTIVE EXAMINAllON (NDE) REll-ABILITY FOR INSERVICE lNSPECTON OF LIGHT WATER MCELROY,W.H.

RE ACTORS.Senvannual ReportApnl 1991-September 1991.

NUREG/CR-3320 V02: LWR PRESSURE VESSEL SURVEILLANCE DO.

LAMBRIGHT,J. SIMETRY IMPROVEMENT PROGRAM. PSF Startup Expenments.

NUREG/CR-5730: RISK EVALUATON FOR A B&W PRESSURIZED MCGARRY,E.D.

WATER REACTOR. EFFECTS OF FIRE PROTECTON SYSTEM AC-TUATON ON SAFETY-RELATED EOU1PMENT. Evaluation Of Genenc NUREG/CR 3320 V02: LWR PRESSURE VESSEL SURVEILLANCE 00- l SIMETRY IMPROVEMENT PROGRAM. PSF Startup Expenments. '1 LANDOW.M. .

MCNA MARA,N.

NUREG/CR 4599 V02 N1: SHORT CRACKS IN PIPING AND PIPING NUREG-0837 V12 NO2: NRC TLD DIRECT RADIATION MONITORING WELDS Senuannual Report, Apni-Septemcer 1991. NETWORK. Progress Report Apro-June 1992.

LARSON.LL MEYER,0.R.

NUREG/CR-6001: AGING ASSESSMENT OF BWR STAND 8Y LIOutD NUREG/CR-6009 V01: DEVELOPING AND ASSESSING ACCIDENT CONTROL SYSTEMS, MANAGEMENT PLANS FOR NUCLEAR POWER PLANTS.Devoicnent Process And Cntena.

NUREG 1465 DAFT FC' ACCOENT SOURCE TERMS FOR LIGHT

• MILLER LA.

WATER NUCLEAR POWER PLANTS. Draft Report For Comment- 1!

NUREG/CR-5305 vot INTEGRATED RISK ASSESSMENT FOR LA-LIANG,H. SALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Risk NUREG/CR-4409 V04 DATA CASE ON DOSE REDL,CTION RE- Uncertainty Evaluation Prograrn (PRUEPK SEARCH PROJECTS FOR NUCLEAR P%ER PLANTS MRER,S.P.

LOPEZ,J.V. NUREG/CR 5910: LOSS OF ESSENTIAL SERVICE WATER IN LWRS -

NUREG/lA-0067: RECIRCULATON SUCTION LARGE BREAK LOCA (Gi-153). Scoping Study.

ANALYSIS OF THE SANTA MAR!A DE GARONA NUCLEAR POWER PLANT USING TRAC 8F1(G1J1). MILSTEAD,W.

- NUREG-0933 S14: A PRORITl2ATON OF GENERIC SAFETY ISSUES.

LU.S.C-NUREG/CR-5864: THEORETCAL AND USER'S MANUAL FOR PC- MINARICK,J.W.

PRAISE.A ProtaabAste Fracture Mecharves Computer Code For Pyr% NUREG/CR-4674 V15: PRECURSORS TO POTENTIAL SEVERE CORE Rehab*ty Analysis- DAMAGE ACCIDENTS: 1991 A STATUS REPORT. Main Report And i

Appendix A.

r LUBBESMEYER.D.

l; NURiG/CR-4674 V16: PRECURSORS TO POTENTIAL SEVERE CORE -

NUREG/LA 0088. NST TEST-ANALYSIS AND NODALIZATION STUO. DAMAGE ACCIDENTS: 1991 A STATUS REDORTAppendices B.C.

IES OF OECD LOFT EXPERIMENT LP-02 6 WITH RELAP5/ MOD 2 And D' CY36 02.

i LUCERO D.A MISKO.C.J<

NUREG/CFi-5443. CORE. CONCRETE INTERACTIONS USING MOLTEN NUREG/CH-5868:

' DEVELOPMENT OF POSITON SENSITIVE PROPOR.

t URANIA WITH ZlRCONIUM ON A _ LIMESTONE CONCRETE TIONAL COUNTERS FOR HOT PARTICLE DETECTION IN LAUNDRY l

BASEMAT,The SURC-1 Excenment. AND PORTAL MONITORS.

l NUREG/CR-5564: CORE-CONCRETE INTERACTONS USING MOLTEN UO(2) WITH ZlRCONIUM ON A BASALTIC BASEMAT,The SURC-2 _ MrTCHELL,0.8. . .

Expenment NUREG/CR 5910: LOSS OF ESSENTIAL SERVICE WATER IN LWRS

. (Gi-150). Scoping Study.

NUREG/CR M RISK EVALUATION FOR A B&W PRESSURIZED MOFFITT,R.

WATER REACTOR. EFFECTS OF FIRE PROTECTION SYSTEM AC- NUREG/CR-5758 V02: F)TNESS FOR DUTY IN THE NUCLEAR POWER l-TVATION ON SAFETY RELATED EQUIPMENT. Evaluation Of Generic INDUSTRY. Annual Summary Of program Performance Reports.CY Issue 57. 1991.

1

Personal Author index 23 MOTE RS.J C. RASMUSON.D M.

NURt G/CR 6779 V01. AGING OF NON POWERCYCLE HEAT EX- NURt G/CP-0124 WORKS 60P DN THE USE OF PHA METHODOLOGY

(>4 ANGERS USED IN NUCLE AR POWER PLANTSOperstmg Enten. FOR TiiE ANMYSIS Of REAC10R EVENTS AND OPERATONAL erre Ard F ailure 6dentrication DATA MURPHYA RASMUSSE N,T.C.

NURE G/CR 6768 V02 IITNE SS FOR DUTY IN THE NUCLE AR POWlR NURE G/CR 6800 NONISOTHERMAL HYDROLOGIC TRANSPORT Ex.

iNDUS1H( Annual Summary Of Program Perioimance Reports CY PERIVENTAL PLAN 1931 FI AVINDR A,M IL NITH L.ML NUREG/CR 4832 Yoh ANALYSIS Or THE LASALLE UNIT 2 WUCLEAR t4UREG/GR $418 TECHN# CAL EVALUATION OF GENERIC ISSUE 113 POWER PLANT: RISK METHODS INTEGRATON Ar4D EVALUATON Ov4AMIC OUAttflCAf TON AND TESTING OF LARGE DOHf Hf- PROGRAM (FiMit P) f aternal t vent $ctswgl Ouantificaton DRAUllC SNUD[l[ R$ NUREG/fM839 METHODS FOR ENTERNAL EVENT SCREENING NUREG/CR 5046 PIPING SYSTEM RESPONSE DURING HOH LivfL QUANTIFICATON RISK METHODS INTEGRATION AND EVALUA-SIMULAlf.D SEISMIC TESTS AT THl. HEISSDAMPt REAKTOR F A. TON PROGRAM (RMTP) ME THODS DENELOPMENT, Cillf Y (SHAM TE ST F ACILfTV) '

REST).

N E CR 2907 Vl& RADOACTIVE MATLAiALS FIELEASED FROM NUCLE AR POWI R PL ANT $ Annuel Report 19t'3 fT N OF XE, C TE DA ND 1 EE CLEAR FUEL UNDER NORMAL AG SEVERE ACCIDENT FR NU l

CONDITONS Users Gude f or Manframa, Worlistation, And Persorel i HOURDAKHSH.H.

ComgMer Applicatone.

NUREG/CR Sete PROOADIUTY AND CONSE')UL NCES Or RAPtO DORON DILUTON IN A PWR A Scopmg StA RETTIG,W.H.

ORTONAD. NUREGrCR 4356 V02 TRAC DF t/MOOLAN ADVANCID DEST.ESTi.

NURE G/CRM01: AGING ASSISSMENT OF F.In STAND 0Y lOUID MATE COMPUTER PROGRAM FOR DOILING WATER REAC'OR AC-CONTROL SYSit MS CfDENT ANALYSIS User's Gude NUREG/CR-4391: TRAC /Df t-MODI MODELS AND CORREL ATONS.

OUVANGA NUREG/CR %85 SE ALING PERrORMANCE Dr DENTONITE AND RICHNE R M.

DENTONITE/ CRUSHED ROCK DOREHOLE PLUGS NURE G/lA 0054 ASSESSMENT Or RELAPS/ MOD 2, CYCLE 36 02.

USING NF PTUN REFLOUDING EXPERIMENT AL DAT A PAGE J D.

NUREG/CR 6410 TECHN: CAL EVALUATON Or GtNERIC ISSui Ita RIDGE LY,J.N.

DYNAMIC OVAttriCATION AND TESTING Of- LARGE HORE HY. NURLG 1465 D. 'T FC. ACCIDLNT BOURCE TERMS FOR ltGHb DRAULIC SNUB 0f R3 WATER NUCLEd POWER PtANTS. Draft Report For Comment PAf4K.J.Y. RIGOS.R.

NUREG/CR4667 V14 ENVIRONMENTALLY ASSISTED CRACKING IN NUREG0933 S14 A PRORIT12ATION Of GENERIC SAFETY ISSUES LIGHT WA1ER REACTORS Semtannual Report October 1991 Masch 1992 ROE SE NE R,W.S.

NUREG/CR 6376 AGING DAT A ANALYSIS AND RISK ASSESSMENT ,

PARKS.M 8- DEVELOPMENT AND DEMONSTRATON STUDY.

NURE G/CP-0120 PROCf EDINGS Or THE rIF TH WOHdSHOP ON CONT AINMENT 8NTEGRITY, held in Wa:Nngton.DC.May 12 14.1992. Otf EA1.

NUREG/CR 5086 EXPERIMENT AL AND ANALYTICAL INVESilGATION P AY NE,A.C.

OF THE SHALLOW 1 LAW EFFECT IN REAC'OR PRESSURE VES-NUREG/CR4832 V0t- ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR $ttg POWER PLANT RISK METHODS INTEGRATION AND EVALUATON PROGRAM (RMIE P1 Summary ROSS S.

NUREG/CR-4032 V02. ANALYSIS OF THE LAS/'LE UNIT 2 NUCLEAR NUREQ/CR 5790' RISK EVALUATON FOR A B&W PRESSURIZED POWER PLANT: RfSK METHODS INTEGRA A AND EVALVATON WATER REACTOR, EFf EJTS OF FIRE PROTLCTON SYSTEM AC-PROGRAM (RM:EP) iningrated Quantricaten And Uncertainty Anaty' TUATION ON SAFETY RELATED COUIPMENTIvaluaten Of Generte NL LG'CR4832 V03 Pt: ANALYSIS OF THE LASALLE UNIT 2 NU-CLFAR POWER PLANT: RISK METHODS INTEGRATON AND EVAL

• ROUHANI,5.L UAtlGN PROGRAM (RMit P) %ernal Events Accr'et So@ence MUREG/CR4;65 V01: TRAC.DFt/ MOD 1: AN ADVANCED DEST-ESil-NU G 3 V03 ANALYSIS Or THE LASALLE UNIT 2 NV. g ygg CLE AR POWER PLANT. RISK Mr,iHODS INTEGRATION AND L' VAL.

NUREG/CR 4356 V07 TRAC DF1/MODt.AN ADVANCED DEST-EST6-UATON PROGRAM (RMit P) Internal Events Accdont Sogaerte MATE COMPUTER PROGRAM FO9 DOILING WAICR REACTOR AC-Quantificator. + rcervices CIDENT ANA YSIS Usva Guede NUREG/CR IM 11: INTEGRATED RfSK ASSESSMEN) FOR LA. NUREG/CR-43[1; TRAC /DF t MOOT MODELS AND CORRELATIONS.

EALLE UN'

• NUCLEAR POWER PLANT ha,.dogy And Fbsk Uncertamty 4 duation Prograrn (PRUE P) pp)ggg,wg J NUREG/CR4667 V14 ENVIRONMENTALLY AS$1STED CRAChlNG IN PP %AN,$8M

.EG S14 A PRORITl2ATON OF GENERIC (LAFETY ISSUES. LIGHT WATER HEACTORS. Sermannual Report,0ctoter 1991. March 1992 PUGHA NUREG/CR $696 AUX 1UARY lEEDWATEfi I 7, TEM RISK-BASED tN- SANECKLJL SPECTION GOOE FOR THE ST. LUCtE bmi 1 NUCLEAR POWER NUREG/CR4667 V14 ENVIRONMENTALLY ASSISTED CRACKING IN GERLRATON STATON LIGHT Y< ATER REACTORS Semennual Report. October 1991. March IDG2.

PUROHIT.A.-

NUREG/CRM667 Vi4 ENVIRONMENTALLY ASSISTED CRACKING IN SCHNUNR,N M.

LIGHT WATER r'EACTORS, Semafvwal ReptvtOctober 1991 March NUREG/CR5673 V02: TRAC PF 1/ MOD 2 CODE MANUAL Usere Gude.

1992.'

SCHWAHN.,5.0.

RAHMANA NUREQ/CR-$868 DEVELOPMENT OF POSITON SENSITIVE PROPOP, NUREG/CRdS99 V02 Nt: SHORT CHACKS IN PIPING AND PIP:NG TONAL COUNTERS FOR HOT PARTICLE DETECTION IN LAUNDRY WELDS Surmannual Report, Apr4 September 1991. AND POR1 AL MONITQR$

F

+ y - -

.--,,w -i-,. ,,,.# ,,_m_mm., ,, m ._~

--~ - ... - . -. -- .- . . ._- --- - - - .-

l 24 Personal Author index i

SCOTT,P. SPANNE R,J C.

NUREGICR.4590 V02 N1: SHORT CRACKS IN PIP!NG AND PIPING NUREG/CR 4469 V13 NONDESTRUCTIVE EKAMINATON (NDE) RELb WELDS Semannual Report Aptd Septemter 1991. ABILITY FOR INSERVICE INSPECTION OF LICHT WATE R Fif ACTORS Serruannual Report Octater 1990-March 1991.

SE C K E R.P.

NURE G/C45819 PROBABluTY AND CONSEQUENCES OF RAPID NUREG/CR4469 V14 NONDE STRUCTRE EKAMINA10N (NDO RELb BORON OtLUTION IN A PWRA Scopmg Study ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER RE#CTORS Sum $ annual Report Apri 1991 September 1991 i SHACK,WJ SPORE.J W NURt G/CR 4667 V14 ENVIRONMENTALLY ASSISTED CRACKING IN IT WATcR REACTORS. Semiannual Report.0cloter 1991 March NUREG/CR 5673 V02 TRAC.PF1/ MOD 2 CODE MANUALUWs Gude.

NUREG/CR 5673 V03- TRAC PF1/ MOD 2 CODE MANUALProgrammer's Guede.

SH AF F E R.C.J. '

NURLG/C44832 V03 Pt ANALYSIS OF THE LASALLE UNIT 2 NU. STEELEA CLE AR POWER PLANT: RISK METHODS INTEGRATION AND EVAL. NUREG/CR $648. PIPING SYSTEM RESPONSE DunlNG HIGRLEVEL UA10N PROGRAM (RM:E Pllnternal Events Accxiant Sequence SIMULATED SEISulC TESTS AT THE HEISSDAMPf REAKTOR FA.

Quantifeebon Mam Report CluTY (SHAM TEST F ACluTY)-

NURLG/CR 4632 V03 P2. ANALYSIS OF THE LASALLE UNIT 2 NV.

CLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVAL, STEINKE.R G.

UAT ON PROGRAM (RM!EP) Internal Events Accdont Sequence NUREG/CR 5673 V02: TRAC-PF1/ MOD 2 CODE MANUALUs#4 Gude.

Quanhfcahon Appendices NUREG/CR-5673 V03- TRAC Pi1/ MOD 2 CODE MANUALProgrammer e Gude.

SHIVER,A.W.

NUREG/CR4832 V02 ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR STE WART.M.A.

POWER PLANT: RISK METHODS INTEGRATON AND EVALUATON NUREG/CR 6009 V02: DEVELOPING AND ASSESSING ACCIDENT PROGRAM (RMIEP) Integrated Quanhficehon And Uncertainty Analy. MANAGEMENT PLANS FOR NUCLEAR POWER PLANTS Evaluahon am Of A Prototype Process.

NL' REG /C46305 VOI: INTEGRATED RISK ASSESSMENT FOR LA.

SALLE UNIT 2 NUCLEAR POWER PLANT Phew ;W And Resit ST OLLER,R E.

Uncertainty Evaluahon Program (PRUEP) NOREG/CR 6859 MODELING THE INFLUENCE OF IRRADIATION TEMPERATURE AND DISPLACEMENT RATE ON RADIATON IN-SHONKA).J. DUCED HARDENING IN FERRITIC STE E LS.

NUREG/CR$8$6 DEVELOPMENT OF POSITON SENSITIVE PROPOR.

TIONAL COUNTERS FOR bdT PARTICLE DETECTON IN LAUNDRY STRAKA.M.

AND PORTAL MONITORS NUREG/CR 6787 V01: TIMING ANALYSIS OF PWR FUEL PIN SHUM.D.K. F AILURES Foal Report Mam Tent Ard Appendices Asl NUREG/C46787 V02- TIMING ANALYWS OF PWR FUEL PIN NUREG/CR $880 EXPERIMENTAL AND ANALYTICAL INVESTIGATON FAILUREST.r.at Report Appenrkas K L OF THE SHALLOWiFLAW EFFECT IN REACTOR PRESSURE VES-SELS STRUCKMEYER,R.

NUREG/CR6008 CONSTRAINT EFFECTS ON FRACTURE TCUGS NUREG-0837 V12 NO2: NPC TLD DIRECT RADIATON MONITORING NESS FOR CIRCUMFERENTIALLY ORIENTED CRALKS IN REA"' NETWORK P/ ogress Report. April. June 1992.

TOR PRESSURE VESSELS SHUMWAY,R W. SYPE T.T NUilEG/CR-4832 V02; ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR NUREG/CR4350 V01: TRAC 'lF1/ MOD 1; AN ADVANCED DEST-ESTb POWER PLANT: RISK METHODS INTEGRATION AND EVALUATON C PROGRAM FOR OWR ACCIDENT PROGRAM (RMIEP). Integrated Ouantifcation And Uncertainty Analy-y3,g y NUREG/CR 4356 V02: TkAC.BF1/ MOD 1.AN ADVANCED BEST ESil- **-

MATE COMPUTER PROGRAM FOR BOILING WATER REACTOR AC. NUREG/C44832 V03 P1: ANALYSIS OF THE LASALLE UNIT 2 NU.

COENT ANALYSIS User's Gude CLEAR FOWER PLANT: RISK METHODS INTEGRAtlON AND EVAL.

NUREG/CR-4391: TRAC /DF1-MOD 1 MODELS AND CORRELATIONS. UATION PROGRAM (RMIEP) internal Events Accident Sequence

- Chaantification.Mam Report SIEF K EN.LJ. NUREG/C44832 V03 P2- ANALYSIS OF THE LASALLE UNIT 2 NU-NUREGICR5787 VOI: TIMING ANALYSIS OF PWR FUEL PIN CLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVAL.

FAIL UNES Fnal Herort Ma n Test And Appendices A.J UATION PROGRAM (RMIEP)Intemal Events Accident Sequence NUREG/CR 5787 V62: TIMING ANALYSIS OF PWR FUEL PIN Quanufcahon. Appendices.

F AILURLS Fmal Report Appendices k L NUREG/CRS305 V01: INTEGRATED filSK ASSESSMENT FOR LA-SALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Arsk .

O EG C 4469 V13 NONDESTRUCTIVE EXAMINATION (NDE) REU-ABluTY .. FOR INSERVICE INSPECTION - OF LIGHT WATER TAYLOR.D.D.

REACTORS Semsennual Report October 1990 March 1991 NUREG/C44356 V01: TRAC BF1/ MOO 1: AN ADVANCED BEST-ESTb NUREG/C44469 V14- NONDESTRUCTIVE EXAMINATON (NDE) RELi' MATE COMPUTER . PROGRAM FO3 - DWR ACCOENT ABluTY FOR INSERVICE INSPECTION OF UGHT WATER ANALYSIS Model Desco'ptM REACTORS Semiannual Report April 1991.Septernter 199 t, NUREG/CR-4356 V02 TR4 LDF1/ MODI AN ADVANCED DEST-ESib geupgog g g, MATE COMPUTER PROGl AM FOR BOluNG WATER REACTOR AC-NUREG/C45443 CORE CONCRETE INTERAC*RONS USING MOLTEN CIDi.NT ANALYSIS User's Guide URANIA- WITH Z'RCONIUM ON A- UMESTONE CONCRETE 74ytog,7 7 NUF E / R h E ONC INTLRACTIONS USING MOLTEN NUREO/C44469 V13: NONDESTRUCTIVE EXAMINATON (NDE) REU-00(2) WITH ZlRCONIUM ON A BASALTIC BASEMAT,The SURC-2 ABluTY - FOR INSERVICE INSPECTION OF LIGHT WATER Expenment REACTORS Serruannual Report Octoter 1994 March 1991.

NUREG/C44469 V14. NONDESTRUCTIVE EXAMINATON {NDE) RELb SlHGER.O.L ABillTY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/C44356 V01. TRAC-DF1/ MOD 1: AN ADVANCED DFST.ESTI. REACTORS Semiannual ReportApnl 1991-September 1991.

MATE COMPUTER PROGRAM -FOR OWR ACCIDENT AN A LYSIS Model Dosenption THEISS,TJ.

NUREG/CR-4356 V02: TRAC-BF1/ MOD 1 AN ADVANCED BEST ESTI. NUREG/CR5886- EXPERIMENTAL AND ANALYTICAL INVESTIGATON MATE COMPU1ER PROGRAM FOR BOtLING WATER REACTOR AC- ' OF THE SHALLOW FLAW EFFLCT IN REACTOR PRESSURE VES-COENT ANALYSIS. Users Guce. SELS.

SOFFERL .

THEOF ANOUS T.G.

NUREGd465 ORFT FC' ACCOENT SOURCE TERMS FOR LIGHT. NUREG/C45854. UNIVE RSAL TREATMENT OF PLUMES AND W ATER NUCLEAR POWER PLANTS. Dra't Report For Comrnent STRESSES FOR. PRESSURIZED THERMAL SHOCK EYALUATONS.

I m.- ,, -. - - _ . _ _ _ - . ,_ , _ __ - . - - , _ - . ~ . - . . . - . . , . - - - - -

Personal Author Index 25 TlCHLER J. NUHI G 14%7 HE 50URCE S A V AIL. ADL L FOH NUCLE AH POWiH NJHL G/CH 2907 V1D HADIOACilVE Mall HiAL$ Ftt Lt A5LD I HOM l't ANT E Mr HGE NCIE S UNDt H THE PHICE ANDI HSON ACT AND NULLi AH POWt H PL ANTS Annua! Hemist 1W9 1HE Holit HT 1 S1 AlIOHD DISAST[H Rfliti AND iMlHGINCY ASSIST ANCE ACT NUHI G/EJ4 %o% HL V!! W AND D(V[ t OrMEN1 OF COMMON NO WE ST R A.C.

ME NCt ATUHE FOH NAMaNG AND LABEUNG SCHIMIS fOH NUNI G/CH 57516 V02 IITNISS F C44 DUTY IN THE NUCLC Au POWEH PHObABillSilC H6% ASSI SSMI NT INDUSTHY Annual Sumnwy Of Prqam Porturmance Hemvis CY V AL10NE N.K.

NUHt G/CH Sa t 9 PHODADIUf f AND CONSLOUfNCES Of IIAPID WHIT E HE AD.D W.

IlOHON DitutON IN A PWit A Scoping btudy NURL G/CH-4832 V02 ANALYSIS of THE LA%At LE UNIT 2 NUCLi AH ViSt LY,W E. POW!H Pt ANT. HISA MllHODS INilGHA10N AND L VALUATON NUHL G/CH %D7 APPfOACHf S f OH AGt-Dt PL NDt NT PROBAbtUS. PHOGHAM WLh Intwated Quantotawn And Uncedainty Ana$

D i i IV TV STU ( N G/CH 48'l2 V03 P1 ANALYSIS Of THE L ASALLI UNIT 2 NU-Cli AH POW 1H PL ANT HISK MLTHODS INTLGHAtlON AND LVAL-V O,T ,V. U AllON PROUMAM (HMLE P) lntornal L vents Austent Swauence NUHf G'CH 44W V13 NONDE S1HUC11'vl I XAMINAllON (NDE ) fitLl- QuantificatKm Main Haport ANUT Y fOH INSl HVlCL INSPE CT K)N OF LK1Hf WATLH NUHf G/CH 4102 V03 P2 ANALYSIS Of 1HE L ASAllt UNif 2 NU-Hi AC10HS 'wmiannual Heyw1 October 19w Mech 1 H1 Cli AH POW [H PLANT Hth ML f HODS INTEGHAf TON AND [V AL+

NUH! G/GH.44% V14 NONDt S11%iUllVL [ x AMrN AtlON (NDI ) ill L l- UATON PHOGHAM (HMil P)Inte nal ivents Acculent Sequence ADillIY $0H INSf HVIC[ INSf*f C1lON Of LIGHT W AT( H Ovantitv;atgm Appervigos.

Hf AClOHS beima/wwal Heport ANd 1991 Septendwr 1W1 NUHL WCH %96 AUnit lAHf f[LDWAl[ H S't Sf! M Hiw HASED IN WICHNE R,R P.

SM ClON GUtDL TOR THE ST tUCil UNil i HUCLi AH POWL H NUNE G/CH %10 [:VAtUATION Of MHIGH FUf L HIUAU:LITY.

GL NL HAitON $1 AliON WILKOWSKI,Q M.

VUt.IN.D S NUHE G/CH 4%9 V02 N1 SHORT CHACkS IN PIPING AND PIPING NUHL G!GH 4409 VD4 DATA CASE ON [OSL RfDUC10N HL- WEL DS Semianrwal Hoport. Apnt Septemter 1991.

St AHCH PHOJL CTS f OH NUCLi AH POWt H PLAN 1S W ADE.N L NUHL G/CH 43% Vol. THAC HF tIMODI AN ADVANCf D filSi iS1b "U O# " "IO # U^ ^ #"^ b ^ b MAIL COMPU11H i fKX1H AM fDH OWH ACCIDI NT DC Vt LOPMi NT AND Di MONSTHAilON STUDY.

AN AL YSIS Mmini (kscroton WOODRUF F S S NUhl G/CH 43% VD2 IhAC DF 1/ MODI AN ADVANCLD BLST E Sf! NUHE G!Ck'N73 V03 TRAC PF1/ MOO 2 CODE MANUAL Pnvammer a MAIL COMPulf H PHOGHAM FOH DOluNG W Af14 HEAC100 AG-CIDE NT AN AL YSis Ow a Guxh UU NUnt 6/CH 4Wi T HAC/DI 1 MOO 1 MODI L $ AND COHu[ t ATIONS NUHf GILH $767 VOI TIM!NG AN AL Y SIS (4 PWH FUIL PIN WU.SL NUHi G/CH 360 VOT, f UI L PtHFOHMANCE ANNUAL HIPORT FOH i AllUHL NUHt G/CH$ Final 57BIHebwrt V2 11 Men Tent Ami MING ANALAg erwisces T IS Of PWH AJ Full PIN IN9 F AllURI S fin 11 H+ twt Apsendg es k L W ALL AC L.D.R= NUHi G/lA 0042 D8SPC HSED f LOW FILM DOtuNG An investigaton Of NUHlG/CH %30 HIGH IN1( GHlf Y SOF TWAHE STANDARDS AND 1he Ponotality To Imswve me Mo@s Imptenented in The NHC Com-GUIDI L INL'S puter Culos For it.e Hetkxuting Ph0se Of The & OCA W AHD,LW. Y AKLE,J L NUH[ G/CH Mov V01. DEVf LOPING AND ASSI SSING ACCIDI NT NUHL G/CH $910 LOSS Of iSSLN1 (L St HVICE W AllH IN LWHS M ANAGE ME NT Pt.ANS FOH NUCLEAH POWlH (G; 1$3) Scoping Study PL ANIS Devekoment Procms And Cntma Y AN.R WAHI,A.O NUHf G/CH SM4 UNIVE RSA' lHLATMINT Of PtVMES AND NuHLG/CH $416 f f CHNIC AL f vat UATON Of GI N( htC ISSUL 113 S1HL SSL S T OH PHI SSUnilED THf HMAL 9400k IVALUAllONS DYNAMIC QUAltFICAfiON AND TESitNG OF LAHGL DOHE HY.

DRAUUC $NUHDLOS Y ARDUMl AN,J.

NUHI G 0525 V01: SAFLGUAHDS SUMMAHY EVENT LIST (S$f t) Pee-WE AVf R W L NHC Ttwough Dec emler 31,19no NUHIG/CH 43% V01 1HAC DF 1/ MOD 1 AN ADVANCID DE St ESil NURE G'OS25 V02 SAF F OU AHDS SUMMAHY PVENT UST MAT [ COMPUT(.H PHOGHAM FOH BWH ACCIDf NT (SMU hn@y 1, IHO DvQ DecerM 31,1991

%N AL 795 Model (Mcnonon NUHf G/Ca l-43% V02 IRAC uf't/ MODI AN ADVANCED DE ST ESTL 2 AW ADZKl,$.A.

MAIL COMPUTE H PROGRAM FOH DO1UNG W ATLH HE ACTOH AC~ NUHE G/CH 5840 F AS1 GRASS A MECHANISitC MODEL FOH THE CIDI NT ANAL YSIS User's Gode PHEDIC110N Of XF, I, CS, IL, B A, AND SH Rtt E ASE "HOM NU-NURI G/CH-4391 T H AC/ Df 1 MODI MODf t S AND COHHEL AllONS CLEAR f ull UNDfH NOHMAL AND SFVthf ACCIDCNT Wt iD( NH AME R,0. COND1110NS UWs Gu*de F or Mainframe. Workstation, And PetsorW NUREG/CH 5316 AGING DATA ANALYSfS AND Hi% ASSESSMENT- Computer Applications DEVt~LOPME NT AND DE MONSlH Af TON STUDY ~

ZHANO.X.J.

WLINSTEIN E. NURE G/CH %67 GRADIENT STUDY Of A LAHGE WILD JOINING NUHf a i442 001 EMERGENCY HEhPONSE HESOURCf S GUOE f or TWO FOHGt D A 506 SHf!LS OF THE MIDLAND HE ACTOH N mw Povun hant Emcegancies VESS[L

1

Subject Index This index was developed from keywords and word strings in titles and abstracts. During this development period there will be some redundancy, which will be removed later when a rea-i sonable thesaurus has boon developed through exporlonce. Suggestions for improvements are welcomo.

1 AEDO NUREG 1242 V02 P01 NRC REVl[W OF ELECTRIC POWER RE.

NUREG 1272 V06 N31. OFFICE f OR ANALYSIS AND LVALUATON OF SEARCH INSTITUTE'S ADVANCED LIGHT WATER FIEACTOR UTILL ,

OP[HATIONAL DAT A 1991 Armuel Report Power Reactors T) REQUIREMENTS DOCUMENT Evolutmary Plant Desgna Chapter  !

NUREG 1272 V06 NO2 Of FICE FOR ANALYSIS AND EVALUATON OF 1 l OPE RATIONAL DAT A 1991 Annual Roport NorweactMs NURtG 1242 V02 P02. NRC REVIEW OF ELECTRIC POWER RE. l SEARCH INSTITUIE'S ADVANCED LIGHT WATER REACTOR UTill.

NU EG/CR 4400 V04 DATA BASE ON DOSE REDUCTON RE. , g SEARCH PROJECTS FOR NUCLEAR POWER PLANTS Adytoory Commmes on Nuclear Weste

1. ^ ^

U IEG/CP 0123 PROCE EDINGS OF THE SECOND NRC/ASME bvM' M 'E N TE r 2 POSlVM ON PUMP AND VALVE TESTING Hold At The Hyatt Regency Hotel.Washuqton.DC, July 21 23. 1992' Aging NUREG/CP-0122 V01: PROCEEDINGS OF THE AGING RESEARCH IN-N bb "^N l NU G/ PO 1 2 i EDINGS OF THE AGING RESEARCH IN.

E Ja y h 1992 NMAT NURE G OO90 V15 NO2 REPORT TO CONGRESS ON ABNORMAL. N G/C ] S M SERVICE WEAR OF SOLENOID-OCCURRENCES ApeJune 1992.

OPERATED VALVES USED IN SAFETY SYSTEM 3 OF NUCLEAR Abstract POWER PLANTS Evaluation Of Morwtorag Methods NUREG 0304 V17 N01 REGULATORY AND TECHSCAL REPORTS NUREG/CH 5378. AGING DATA ANALYSIS AND RISK ASSES $ MENT ~

A N X JOURNAL) Comsdiatton f or Fnst Ouarter L E A EEM iR T S UDY p

TIC SAFETY ASSESSMENTS WITH EMPHASIS ON PRIOH!Tl2ATiON '

Accident Analysle AND 6ENSITIVITY STUDtES NURIG/CR43% Vol. TRAC-DFt/ MOO 1 AN ADVANCtD DE ST-ESil. NUREG/CR 5700 AGING ASSESSMENT OF REACTOR INSTRUMEN.

MATE COMPUTE R PROGRAM FOR OWR ACCIDENT T ATION AND PROTECTION SYSTEM COMPONENTS.Agng Rntated AN ALYSIS Modal Descr phon Operating Emperences f4UREG/CR-4%6 V02: TRAC-DFt/MOOf.AN ADVANCED BEST ESil- NUREG/GR 5772 V01: AGING. CONDITION MONITORING. AND LOSS-MATE COMPUTER PROGRAM FOR DOILING WATER REACTOR AC. OF COOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E ELECTR6 CAL CIDENT ANALYSIS User's Gunte CABLES Crosslinked Poho4ehn Cat >les NUREG/CR 5719 VOI: AGING OF NON POWf R-CYCLE HEAT Ex-Acc6 dent Management CHANGERS USED IN NUCLEAR POWER PL ANTSOperehng Emper6 NUREG/CR 6009 V01' DEVELOPING AND ASSESSING ACCIDENT ente And Feelure identshcahon MANAGE ME NT PLANS FOR NUCLEAR POWER NUREG/CR-6001: AGING ASSESSMENT OF DWR STANDDY LIQUID PL AN1S Development Process And Cnloria CONTROL SYSTEMS NURE G/CR-6009 V02 DEVELOPING AND ASSES $1NG ACCIDENT MANAGEMENT PLANS FOR NUCLE AR POWER PLANTS Evalunhon Aging Mitigation Of A Prototype Process NUREG-1377 R03 NRC RESEARCH PROGRAM ON PLANT AGING LISTING AND SUMMARIES Or REPORTS ISSUED THROUGH JULY Acc6 dent Sequence 1992.

NUREG/CR4674 V15. PRECURSOR $ TO POTENTIAL SEVERE CORE DAMAGE ACCIDENTS 1991 A STATUS REPORT Main Report And Airt>orne Effluent Appervsm A. NUREG/CR 2007 V10. RADCACTIVE MATERIALS RELEASED FROM NUREG/CR 4674 V16 PRECURSORS TO POTENTIAL SEVERE CORE NUCLEAR POWER PLANTS. Annual Report 1960 DAMAGE ACCIDENTS 1991 A STATUS REPORT.Appendicos B C, And D Annual Reposi NUREG-1145 V08. U S. NUCLEAR REGULATORY COMMISSON b1 Accident Sequence Analysis ANNUAL REPORT.

NUREG/CR 4632 V01 ANALYSIS OF THE LASALLE UNti 2 NUCLEAR NUR(G 1470 V01; CHIEF FINANCIAL OF FICEf4 S ANNUAL REPORT -

POWER PLANT: MISK METHODS INTEGRATON ANO EVALVAtlON 1992.

PROGRAM (RMIEP) Summary. NUREG/CR 3950 V07; FUEL PERFORMANCE ANNUAL REPORT FOR Accident Sequence Quantifwatton "

NUREG/CR-4832 V03 Pt: ANALYSIS OF THE LASALLE UNIT 2 NU- Austilary Feedwater System CLEAR POWER PLANT. RISK METHODS INTEGRATON AND EVAL- NUREG/CR $896. AUXILIARY FEEDWATER SYSTEM Risk DASED IN-UATON PROGRAM (RMT P) Internal Events Acadent Sequence SPECTION GUIDE FOR THE ST. LUCIE UNIT 1 NUCLEAR POWER Ouanhficabon Main Report GENERATION STATON NUREG/CH-4832 V03 P2: ANALYSIS OF THE LASALLE UNIT 2 NU.

CLE AR POWER PLANT. RISK METHODS INTEGRATON AND EVAL- BWR UATON PROGRAM (RMiEP)lnternal Events Accadent - Sequence NUREG/CA 43% V01: TRAC DFt/ MOO 1: AN ADVANCED DEST ESil-QuantitcattortAppendtes MATE COMPUTER PROGRAM FOR DWR ACCIDENT ANALYSIS Model Descnpbon.

Advanced Eight Water Reactor NUREG/CR4356 V02 TRAC-DF1/ MODI.AN ADVANOED DIST ESil-NUREG 1242 V01' NRC REVIEW OF ELECTRIC POWER RESEARCH MATE COMPUTER PROGRAM FOR DOILING WATER REACTOR AC-tNSTITUTE'S ADVANCED L6GHT WATER REACTOR UTILITY RE- CIDENT ANALYSIS User's Guede OUIREMENTS DOCUMENT Program Sumrnary NUREG/CR-4391: TRAC /DF1 MOD 1 MODELS AND CORRELATONS-27

g .-

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}

i i  :

i 28 Subject Index 4 l

NUREGICR 4001 AGING ASSESSMENT OF BWR STAND 8Y UOUID Core Concrete j CONTMOL SYSTEMS NUREQ/CR4443 CORECONCAETE INTERACTONS USING MOLTEN NUREG/CR v>03 DENSITOWAVE tNSTADIUTIES IN BOluNG WATER URANIA WilH ZlRCONIUM ON A UMESTONE CONCRETE  !

REACTORS- BASEMAT.The SURC 1 E apenment NUREGICR M64 CORE CCNCRETE INTERACTONS USING MOLTEN

^ ^ 0^

N G/ 4 CORE CONCRETE iNTERACTONS USING MOLTEN **"I UOm WITH ZlRCONIUM ON A BASALTIC BASEMAT.The SURC-2 Esponment Crack gm , NUREG/CR 5872: ORNOZL: A FINITE-ELEMENT MESH GENERATOR FOR NO22LE CYUNDER IfdTERSECTIONS CONTAINING INNER-NUREG/CP 5443 CORECONCRETE INTERACTONS USING MOLTEN CRACK URANIA WITH ZlRCONIUM ON A UMLSTONE CONCRETE g ihNE f

BASE MAT Tte SURC i tupenment NESS FOR CIRCUMFERENTMLLY ORIENTED CRACKS IN REAC-Bentonite 10A PRESSURE VESSELS NURE G/CR %85 SEAltNG PERFORMANCE OF BENTONfTE AND BEN 10 NITE /CRUGHED ROCK BOREHOLE PLUGS. Crack owth Bomng Water Reactor LIGHT WATER RE ACTORS Somaannual Report. October 1991. March NUREG/CR-4356 VOI: TRAC BF1/ MOD 1 AN ADVANCED BEST.ESil- 1992, MATE COMPUTER PROGRAM FOR BWR ACCIDE NT ANALYSIS Modet Detuipton Decommisaloning NUREG/CR 4356 V02. TRAGBFt/ MODI.AN ADVANCED BFST-ESil- NUREG/CR-5849 DAF FC: MANUAL FOR CONDUCTING RADIOLOGI-MATE COMPUTER PROGRAM FOR BOluNG WATER REACTOR AC- CAL SURVEYS IN SUPPORT OF LICLNSE TERMINATION Draft CIDE NT ANALYSIS User's Guwte Report For Comment NUHEG/CR-4391: TRAC /BF1 MOD 1 MODELS AND CORRELATONS.

NURE G/CR 4001: AGING ASSESSMENT OF BWR STAND 8Y LIQUlO Decontaminst60n CONTROL SYSTEMS NUREG 1442 R01; EMERGENCY RESPONSE RESOURCES GUIDEfor NUREG/CR-6003. DENGITY WAVE INSTABlWTIES IN BOluNG WATER Nucioat Power Plant Emerganoes REACTORS Deep Croch

• NUREG/CR 5793- A COMPARISON OF ANALYSIS METHODOLOGIES NUREG/CR %85 SEAllNG PFRFORMANCE OF BENTONITE AND FOR PREDICTING CLEAVAGE ARREST OF A DEEP CRACK IN A RE-ACTOR PRESSURE VESSEL SUBJECTED TO PRESW sRIZED THER, NL E 5 7 i TV NSELY WELDED MAL SHOCK LOADING CONDITIONS, yur g y Doron Dilution Density Wave NJREG/CR'5810 PROBABruTY AND CONSEQUENCES OF RAPID NUREG/CR4001 DENSITY-WAVE INSTADIUTIES IN BOluNG WATER DORON DILUTON IN A PWR A Scopmg Study REACTORS.

Charpy V Notch Design Basis Accident NUREGICR 5891: ACCELERATED 1RRADtATON TEST OF GUNDREM- NUREG 1465 DRFT FC: ACCOENT SOURCE TERMS FOR UGHT.

MtNGEN REnCTOR VESSEL TREPAN MATERIAL WATER NUCLEAR POWER PLANTS. Draft Report for Comment Chief Financial Officers Act Dose Reduction NUREG-1470 V01: CHIEF FINANCML OFFICER'S ANNUAL REPORT - NUREG/CR-4409 V04 DATA BASE ON DOSE REDUCTION RE.

1902. SEARCH PROJECTS FOR NUCLEAR POWER PLANTS.

Cladding Youghness Doelmetry NUREG/CR 5867: GRADIENT STUDY OF A LARGE WELD JOINING NUREG/CR 3320 Vfs2. LWR PRESSURE VESSEL SURVEILLANCE DO~

TWO FORGED A 508 SHELL3 OF 1HE MIDLAND REACTOR SIMETRY lMPROVEMENT PROGRAM PSF Startup Expenments, VESSEL i Embntitement Class it Electricel Cable NUREG/CR-5891: ACCELERATED IRRADIATON TEST OF QUNDREM-NUREG/CR 5772 V01. AGtNG. CONDITION MONITORING. AND LOSS- MINGEN REACTOR VESSEL TREPAN MATERIAL OFCOOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E ELECTRICAL l CADLES Crosshnked Polyoletm Cet9es. Emergency Response l NUREG-1442 R01: EMERGENCY RESPONSE RESOURCES GUIDE For Cleav80s Airest N NUREGICR 5793. A COMPARISON OF ANAL *' SIS METHODOLOGIES gg jarf57 'n 1 R ILABLE FOR NUCLEAR POWER FOR PREDICTING CLEAVAGE ARREST OF A DEEP CRACK IN A RE' PLANT EMERGENCIES UNDER THE PRICE-ANDERSON ACT AND ACTOR PRESSURE VESSEL SUEUECTED TO PRESSURIZED THER- THE ROBERT T, STAFFORD DISASTER REUEF AND EMERGENCY MAWOCK LOADING CONDITONS ASSISTANCE ACT.

Component Reliability EnWeement Acuon NURE G/CR 5587: APPROACHES FOR AGE-DEPENDENT PROBABluS-TIC SAFETY ASSESSMENTS WITH EMPHASIS ON PRORIT12ATION NUREG 0940 Vit NO2' ENFORCEMENT ACTONS- SIGNIFICANT AC.

TONS RESOLVED Quarterty Progress Report, April-June 1992, AND SENSITIVITY STUDIES.

- Containment Integrtty Engineering Practice NUREG/CR-5930; HIGH INTEGRITY SOFTWARE STANDARDS AND NUREG/CP-0120: PROCEEDINGS OF THE FIFTH WORKSHOP ON CONTAINMENT INTEGRITY. He6d in WasNngton.DC,May 12 14,1992. GUOEUNES.

Core Damage Essential Service Water NUREG/CR-4674 V15 PRECURSORS TO POTENTIAL SEVERE CORE NUREGICR4910: LOOS OF ESSEN1ML SERVICE WATER IN LWAS DAMAGE ACCOENTS: 1991 A STATUS REPORT. Man Report And (Gt 153). Scoping Study.

NU C 4674 V18. PRECURSORS TO POTENTIAL SEVERE CORE Event Tree l DAMAGE ACCIDENTS 1941 A STATUS REPORT, Appendices B.C. NUREG/CR 4874 V15: PRECURSORS TO POTENTML SEVERE CORE And D. DAMAGE ACCIDENTS.1991 A STATUS REPORT. Man Report And A mendw A-Core Mettdown NUREG/CR-4674 V16. PRECURSORS 10 POTENTIAL SEVERE CORE NUREG 1465 DRFT ?C ACCOENT SOURCE TERMS FOR UGHT- DAMAGE ACCOENTS: 1991 A STATUS REPORTAppendices B.C.

WATER HUCLEAR POWER PLANTS Dratt Report For Comment. And D. .

i s

Wk1

Subject index 29 E stern.i E vent f uei NURLG/CH 41132 V07 ANALYblS OF 1HE LASALLE UNIT F NUCLE AR NUREG/CH 5810 EVALUATON OF MHYGR FUEL hf LIABlUTY-POWE R PLANY: HISK METHODS INTEORATION AND LVALUA10N PROGRAM (HulE P1 E v1ornal Event Scofva Quantite abon fuel Performance NUHt G/CH 4839 ML THODS F OR t # TL HNAL L VE NT SCHE ENING NUNEG/CH 3950 V0f IUEL PtHf 0HMANCC ANNUAL HEPORT FOH QUAN11 FICA TON RISK MI THODS INTEGRAT80N AND E VALVA- gm TON fHOGRAM (HM4 P) Ml1 HODS DEVELOPMENT.

FASTGRASS Fuel P6n NUHEG/CH iB40 f A$f GHASS A MICHANtSTIC MODEL FOH THE NURE Q/C4 5767 V01, 11 MING ANALYSTS OF PWR I Uf L PIN PHEDCTON OF XE I, CS, TE, DA, AND S4 HE LEASE I HOM NU- T All UHf S f mal Report Mam f eat And Appendees A J CLIAR # DEL UNDf.H NOHMAL AND SEVf HE-ACCIDt Nf NUHEG/CH 5767 V02 11 MING ANALYSIS OF PWR FUEL PIN CONDl! IONS Utar s Gukie f cw Mainframe, workstatior( And Personal F AILUNES F mal Appcwt Apperuhces k C Cumputer Aptkatoons Gate Valve Failure Mode NUMEG/CH $648 PirlNG SYSit M HESPONSE DUR!NG HIGH LEVEL NUHFG/GR 5810 (VALUATION Of MHtGR l'Ull AlllABilliY SIMUL ATED SEl5MIC flSTS AT THE HEISSDAMPFHC AKf0R FA-f ault Displacement Hazard CillTY (SHAM 1ESI FAClllTY)

NUHEG 1451 ST AF F 1( CHNICAL POSITION ON INVf SilGATIONS TO Genwsuon 3 DeWgn IDI Nilf Y I AULT lhSPLACT ME NT HAZANDS AND SEISMIC HAZ-ARD9 AT A G60LOGC HEPOSi1OHf. NUHf G/CH 6010. HISTORY AND CURRENT STATUS OF GiNERA.

1ON 3 THEHMAL SLt EVE S IN Wt $1tNGHOUSE NUCLf AR POWER Terrific isteel PLANTS NUMEG/CH48% MODf tlNG THE INFlut NCE OF IRH ADI ATION TE MP( HATURE AND (hSPLACE MENT HATE ON RADIATON IN Generator DUCE D HAHDENING IN F EHRiflC SIEELS. NU4f G/CA 6877 OHNOll: A FINIf Ef tWENT MESH GEN!RATOR f 0R NOZILE CYLINDin INf f RSECTONS CONTAINING INN (R.

b U G A 0042. DISDT RSED FLOW FILM DOILfNG An investvaahon Of The Possilahty To improve The Models implomonted in t he NHC Co'* Generic laeue 073 potes Codes f or the Refloorling Phase Of the LOCA NUnf G/CH M10. HISTORY AND CUnHENT STATUS OF GENERA.

Finas Design Approvst ilON 3 THE RMAL SLEEVES IN WLSTINGHOUSE NUClLAR POWEH NUHEQ 1242 V01: NHC HEVIEW OF ELEC1HtC FOWEH HLSE AHCH PLANTS INSTITUTE'S ADVANCED ttGHT WATER HE ACTOR UTitiTY DE-OtDRE MENTS DOCUMI NT Plouram Summww Generic issue IIS NURLG 1242 V02 POI. NHC HI vil W OF (Li C1RIC POW (R Hb NUhlG/CH $4164 TECHNICAL EVALUATION OF GENEMIC IS$Uti 113 SE ARCH INSilIUTt"S ADVANCED LIGHT WATER HE ACTOR Uilla. DYNAM C QUALIFICAlON AND TESilNG Or (AnGE DOHE HY-TY ACQUlHL MENTS DOCUMENT. Evolutamary IMant Designa Chapter DRAULIC SNultDE R$

1 NUREG 1242 V02 FW- NRC REVIEW OF EI' OTRIC POWER HE- Generic $afety issue St AHCH INSilIUff S ADVANCED LIG6tT WATEA RE ACTOR UllLI- NUREG 0933 S14. A Pntonlil/ATON OF Of NEHIC SAFITY ISSUCS.

TV HE OUlHE MI NTS DOCUMENT. Evolutionary Plant Designe Charlers 213 Geologic flopository Financial Management NUREG 1451: STAFF TICHNICAL POSITION ON INVE SilGATlONS TO NURE G '470 VOI: CHTF f lNANCIAL OF FICER S ANNUAL HIPORT - IDENTIFY F AULT DISPLACEMENT HAZANDS AND SEISMC HAZ.

ARDS AT A GEOLOGIC Rf POSif 0HY, 1992.

fire ProtecHon System O'odient Study NURE G/CH-5790 RISK EVALUATON f on A BAW PRESSUH1 ZED NUREG/CR 586h GRADIENT STUDY OF A LARGE W"LD JOINING WAltH REACTOR (TT ECTS OF FIRE PROTECTION SYSTf M AC. TWO FORGED A 608 SHELLS OF THE s5*ANL REACTOR TUATION ON SAFETY HELATED EQUIPMENT Evaluauon Of Genene VE SSEL issus $7, Hardening Flee 6on Gas NUMEG/CA S859 MODELING THE INrLUENCE OF IRRADIATION NUHEQ'CR 5840 F ASYGRASS A MECHANISTIC MODEL FOR THE T[ Mpg:RATURE AND DISPLACEMENT f4 ATE ON HAD!ATON IN-PHEDCTION C# XE. I, CS. TE, DA, AND SR AELEASE FROM NU* DUCED HARDEN 4NG IN F ERRiflC STI CLS.

CICAR FUEL UNDE R NORMAL AND SLVERE-ACCIDENT CONDITIONS User's Gunto For Mamtrame, Workstahon, Arwi Personal Heat Exchanger Computer Appheabons NUREG/CR $779 Vot, AGING OF NON POWER CYCt E HEAT M Fitness For Duty CHANGERS USED IN NUCL.EAH POWER Pt ANTS Operatmg Entwn-NUHLGICR 5756 Vol FITNESS FOR DUTY IN THE NUCt E AH POWER ence And Fadure suentkahort STRY Annual Summary Of Program Performance HeportsCY High integrity NUREG/CH4930 HIGH INTEGRITY SOFTWARE STANDAHOS AND Ftecture Mechanics GUIDEllNES.

NURI G/CH 4599 V02 Nt: SHORT CHACKS IN PAPING AND PIPING WEL DS Semiannual Report1 Apsd Septemtwar 1901 High-Level Rad 60 active Weste NURE G/CH-5864 THEOHETICAL AND USER'S MANUAL FOR PC- NUHEG 1423 V03 A COMPILATON OF REFORTS OF THE ADVISORY PRAISE A Probatphshc Fracture Muhames Computer Code For Piping COMMITTEE ON NUCLEAR WASTEJuly 1991 June 1992 Rehabehty Anahsis NUREG-145t: STAFF TECHNCAL POSITION ON INVESTIGATONS TO IDENTIFY FAUL T DISPLACEMENT HAZARDS AND SOSMC HAZ-

^ ^

U EG C 58 f3PERIME NTAL AND ANALYTCAL INVESTIGATON OF THE SHALLOW f LAW EFFECT IN REACTOR PRESSURE VES- High Level Radioactive WasteDisposal NU /CR 6008. CON 3TRAINT EFFICTS ON FRACTURE TOUGH NUREG/CR4880. NONISO1HERMAL HVDROLOGC i AAN3 PORT EX-NsSS FOR CIRCUMFERENiiAL LY ORIE NTED CRAOkS IN REAC- PE RIME NT AL PLAN TOR PRESSURE VESSELS g,, p,79,c,,

Fractured Rock NUHEQ/CR St -1 DEVELOFMENT OF POSITION SENSITIVE PROPOR-NUAEG/CR48fl0 NONISOTHt AMAL HYDAOLOGIC TRANSPOHT EX- TIONAt COUNTERS FOR HOT PARTCLE E,ETECTION IN LAUNDRY PEH31ENT AL PLAN ANO PORTAL MON: TORS

/.

w

i 30 Subject index  :

Human Factor EWR NUREG/CR 0009 V01: DEVELOPING AND ASSESSING ACCOENT NUREG-1242 V01: NRC REVIEW OF ELECTRIC POWER RESEARCH MANAGEMENT PLANS FOR NUCL E AR POWER INSTITUTE'S ADVANCED LOHT WATER RE ACTOR UllLITY RE-PLANTS Dervedpment Process And Criter$a OUIREMENTS DOCUMENT Propam Summary NURLG/CR 6009 V02 DEVELOPING AND ASSES $1NG ACCIDENT NUREG 1242 V02 Pot; NRC REVIEW OF ELECTRC POWER RE-MANAGEMENT PLANS FOR NUCLEAR POWER PLANIS Evaluaton SEARCH INSTITUTE S ADVANCED LIGHT Ws.TER REACTOR UTill.

Of A Prototype Process TY REOUtREMENTS DOCUMENT, Evolutonary Piant Desgns Chapter 1

< Hydrautic Snutet er NUREG-124 V02 P02- NRC REVIEW OF ELECTRIC POWER RC-NUREG/CH 5416. TECHNICAL EVALUATON OF GENERIC ISSUE 113 SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR UTILI-DYNAMIC OVAliFICATON AND TESTING OF LARGE DORE HY- TV REQUIRE MENTS DOCUMENT. Evolutonary Plant DRAUllC SNUDDERS Desons Chapters 213 Hydrodynarn6c NUREG/CR 3320 Vo2 LWR PRESSURE VESSEL SURVEILLANCE DO-SIMETRY IMPROVEMENT PROGRAM PSF Startup Erpenments.

NUR( G/CR-5673 V02: TRAC PFI/ MOD 2 CODE MANUAL Users Gude. .

NURIG/CR 5673 VD) TRAG PF1/ MOD 2 CODE MANUAL Programmer's NUREG/CR.4469 V13: NONDESTRUCTIVE EXAMINATON (NDE) REU-Gude ABILITY FOR INSERVICE INSPECTON OF LtGHT WATER ,

RE ACTORS Sermannual Aeport October 1990 March 1991.

Hydrolopic Transport NUREG/CR-4469 V14 NONDESTRUCTIVE EXAMINATION (NDE) REU-NUREG/CR.5BBQ NON!SOTHEhMAL HYDROLOGIC TRANSPORT EL ABILITY FOR INSERVICE INSrECTON OF LIGHT WATER PERIMENTAL PLAN. RE ACTORS Semannual Fleport April 1991 Septemter 1991 NUREG/CR-4067 Vie: ENVIRONMENTALLY AS$1STED CRACKING IN EG 4 1ESOURCES AVAILABLE FOR NLOLEAR POWTR gg PLANT EMERGENCtES UNDER THE PRICE ANDERSON ACT AND NUREG/CR4910. LOSS OF ESSENTIAL SERVICE WATER IN LWRS THE RODERT T. STAFFORD DISASTER REllt F AND EMERGENCY (Gl.1$3) Scoping Study ASSIST ANCE ACT.

Earge Bore in9estion Pathwsy NUREG/CR-5416: TECHNICAL EVALUATION OF GENERC ISSUE 113 NUREG 1442 Rot: EMERGENCY RESPONSE RESOURCES GUIDE Fo' DYNAMIC OUALIFICATON AND TESTING OF EARGF DORE HY-Nuclear Power Plant Erregences. DRAULIC SNUBBERS.

Inservice inspect on AB L T S RV SE Id F H T NURLG R S 68 DEVELOPMENT OF POSITON SENSITIVE PROPOR-RFACTORS Semeannual Report Or:tnter 1990 March 1991. TONAL COUNTERS FOR HOT PARTCLE DETECTON IN LAUNDRY NUREG/CR-4469 V14. NONDESTRUCTIVE EXAMINATON (NDE) RELj. AND PORTAL MONITORS.

ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER RE ACTORS Semiannual ReportApnl 1991 Septemta 1991.

"f V3$ N05 NUCLEAR REGULATORY COMMISGON IS-Inservice Testing SUANCES FOR MAY 1992 Pages 169-201 o NUREG/CP 0123 PROCEEDINGS OF THE SECOND NRC/ASME SyM. NUREG 0750 V35 N06: NUCLEAR Rt.GULATORY COMMISSION IS-POSIUM ON PUMP AND VALVE TESTING Held At The Hyatt Regency SUANCES FOR JUNE 1992 Page 205-200.

Hotel, Washington.DC, July 21 23, 1992, integrated Ouent!fication NUREG/CR-5649 DAF FC: MANUAL FOR CONDUCTING RADIOLOGI-NUREG/CR-4832 V02: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR CAL SURVEYS IN SUPPORT OF LICENSE TERMINATION Draft

. POWER PLANT: AISK METHODS INTEGRATION AND EVALVATON Report For Comment PROGRAM (RMIEP). Integrated QuantAcaton And Uncertanty Anaty-sis Ught Watef Reactor ..

NUREG/CR-3320 V0E LWR PRESSURE VESSEL SURVEILLANCE DO-Integrated Risti Assessment $1tETRY IMPROVEMENT PROGRAM PSF Startup Expenments.

NUREG/CR-5305 V01: INTEGRATED RISK ASSESSMENT FOR LA- NUREG/CR-4469 V13 NONDESTRUCTIVE EXAMINATON (NDE) REll-SALLE UNIT 2 NUCLEAR POWER PLANT PtNnomenology And Risk ABILITY FOR INSERVICE INSPECTION OF LIGHT WATER Uncertamty Evaluat on Program (PRUEPl REACTORS Semiannual Report October 1990-March 1991.

NUREG/CR-4469 V14. NONDESTRUCTIVE EXAMINATION (NDE) REll-I"I"*' I "I ABluTY FOR INSERVICE INSPECTION OF LIGHT WATER NUREG/CR-4632 V03 Pt: ANALYSIS OF THE LASALLE UNIT 2 NW RE ACTORS Semianreal Report April 1991. September 1991.

CLEAR POWER PLANT: RISK ME1 HODS INTEGRATION AND EVAL

• NUREG/CR 4667 Vie: ENVIRONMENTALLY ASSISTED CRACKING IN UATON PROGRAM (RMIEP). Internal Events Accident Sequence Quantifcation Main Report L,IGHT pgp WATER REAC' ORS Semiannual ReportOctote 1991 + March NUREG/CR 4832 V03 P2. ANALYSIS OF THE LASALLE UNIT 2 NU. NUREG/Cfw 310- LOSS OF ESSENTIAL SERytCE WATER IN LWRS CLEAR POWER PLANT: RISK METHODS INTEGRATON AND EVAL-UATION PROGRAM (RMIEP) internal Events Accident Sequence I ""*"9 8'*

Quantif<aton Append <es (ogg.Of Coolant Accident NUREG/CR 6772 VOI: AGING, CONDITION MONITORING, AND LOSS-irradiation NUREG/CR 5859. MODELING THE INFLUENCE OF IRRADIATION OF COOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E ELECTRICAL l

"^ ""^'^ NU EG $ 87 Y1 G LYSIS OF. PWR FUEL PIN .

HARDEN N N A T ST L S~ F AILURES Final Report. Main Text And Appendest A J.

4 erradiation Test NUREG/CR 5787 V02- TIMING ANALYSIS OF PWR FUEL PIN NUREG/CR 589t; ACCELERATED 1RRADIATON TEST OF GUNDREM. FAILURES Final Report Appereces KL a

MINGEN REACTOR VESSEL TREPAN MATERIAL.

-Low-level Radioactive Waste LOCA NUREG-1423 V03. A COMPILATION OF REPORTS OF THE ADVISORY ~

NUREG/LA 004E DISPERSED FLOW FILM BOillNG.An invest. gabon Of COMMITTEE ON NUCLEAR WASTEJuly 1991 June 1992.

, The Possiteltty To improve The Models implemented in The NRC Com-puter Cot 1r1 For The Refloodmo Phase Of The LOCA. WHTGR 1 NUREG/ LAO 6?: RECIRCULATION SUCTION LARGE DREAK LOCA NUREG/CP&10: EVALUATON OF MHTGR FUEL RELIABILITY.

ANALYSIS OF THE SANTA MARIA DE GARONA NUCLEAR POWER

_4 Mechantsuc Model PLANT USING TRAC BF1(GtJt)

- NUREG/CR-5847 FASTORASS A MECHANISTIC MODEL FOR THE .

< LOFT Esperiment LP42-4 PREDICTON OF XE, L CS, TE, BA. AND SR RELEASE FROM NU-NUREG/lA 0088. POST TEST ANALYSIS AND NODAll2ATION STUD. CLEAR FUEL UNDER NORMAL AND SEVERE-ACCIDENT

, . IES OF OECD LOFT EXPERIMENT LP-02-6 WITH RELAPS/ MOD 2 CONDITONS User's Guide For Mainframe. Workstation And Personal CY36 02- Computer Apphcations.

1

- . , - , ,wn, .-v.. , , . - , , . , _ . , , , , , .w,, ,,,,,,,w,.n.,, , - , , , _ , , _ ,.,,,-,,+.,,.w... - . , . , , , , , , , , , , , , ..,,.-,.--m , , , , , , + . , , , . , ~ - , -

Subject inden 31 NURf G/CR 4837 V02 ANAL YSl$ OF 1HE LASALLE UNil 2 NUCLE AR NtPTON A%E SSMl NT OF HI LAPS /MODL CYCLE 3602, POWLR PLANT. RISK METHODS IN11GRAllON AND F VALUATION NUHLGnA tc54 PROGRAM (HMiEP) Inteysted Quanbhcaton And Uncertainty Ar aty-USING hiPTUN RU LOODING ( APE RIMENT AL DAT A sis NUHEG/CR 4832 V03 P1. ANALYSIS OF THE LASAt1i UNif 2 NU-Namtry And Laboung Scheme CLEAR POWE R PLANf HISK Ml f HODS IN1LGRAflON AND LVAL-HUH [G/Cl4 %35 HEVitW AND DEVELOPME NT OF COMMON NO- PROGRAM (HMit P)1ntemal ( vents Accdont Se'.3uence M(NCt AT UHE FOR NAMING AND LAL4 LING SCHEMtS FORU A10,Nos.r3,caton Mam Npcwt PHOUAbtuSTC HISK AS$tSSMt N1. NUHLG/CR.4632 V03 P2. ANALY5iS Of " THE^LASALLE UNIT 2 NU-U AU IN ##"I "

r Sws RL Vif W AND DrVELOPMt N1 Or COMMON NO UA EN I D I"#"' I"""

VINCLATUHE FOR NAMi4G AND 1.ADluNG SCHE f 4LS $ 0R g f V[hf[AlYSIS OF THE LASALLE UNIT 2 NUCLEAR

$H0H ADLLISTIC AtSK AS5LSLMf N1 POWE H PLAN 1 Ri% METHODS INT!GRATON AND E VALUATON PHOGRAM @@lPi f sterna' [ent Scoging Oaantif(aton Nondestructive t amminat6on NUHtG/CH 4409 V13 NONDE STRUCilVE E RAMINATON (NDE) NUHE Ll. 4819 MfiHODS FOR L Xi[HNAL EVENT SCRLE NING RtG/CR Ahlll1Y FOR INSL HVtCE INSPtG10N Or tnGHT WAf tR QUANTIFICAliON RISK METHODS INTEGRATION AND E VALUA.

Rt ACTORS Semanque Nmrt (Moter 1990 March 1991 lON PROGRAM (f4Mi[P) METHODS DEVELOPME ASSESSMLNT NT.

f 0R LA NUHi G/CH-4469 Vtd NONDLS1HUC11VE L XAMiNATION (NDf) NUHtG/CH REU-UNIT SMS V01. INTEGRA1LD RISit SALLE 2 NUCLE AR POWE R PLANT Phenomenology And R*k AfilllT Y FOR IN5t HVlCE IN5P( C10N Of itGHT WATIR HE ACTORS Semiannual Npret Apnl 1991 Septemtw 1991. Uncertainty i valuation Pwyam (PHUE P)

NUHEG/CR $378. AGING DAT A ANALYSIS AND RISK AS$tSSMENT-Nonisothermal Ilow DEV! LOPMI.N1 AND Di MONSTRAtlON STUDY.

NUHf G/CH Mn0 NONISOTHf RMAL HYDHOtOGIC TRANSPORT NUHt EX-G/CR M96 AUKlWARY f[lDWAT[R SYSTEM RtSK-84stD IN-M fuMLN1 AL PLAN LPECTON GutDL FOH 1HE $1 LUCtf UNif 1 NUCLEAR POWE4 Gt NtHAllON 51 ATON.

Norste-Cyhnder NUrdG/CR 09V5. filVi[W AND DEVELOPMENT OF COMMON NO-NUitlG/C94872 OHNO?L A flN11[ EllME NT M15H GEN (HATOR MI NCLATUHE TOH NAMING AND LABEUNG SCHEMES FOR IDH NOlllt CYuNDin 14TE R5(CitONS CONI AIN)NG JNNER- PHORADtUSilC Al% ASSESSMENT.

COHNE R Cf 4ACK5 NURLG'CR 5910 LO59 OF [S$[NTIAL SCRVICE WAttR IN LWOS O

L f G /lA40H9 50St.TI ST AN ALYLtS AND NODAlt/ ATON $1UD- pwn it s Or OECD LOf 1 LW0$UMENT LP-024 WITH HLt AP5/ MOD NURE 2 G/CR 67B7 V01 TlutNG ANALYSIS OF PWR FULL PIN CYW O2 F AILUMLS hnal AcWasn Tent And Appeedces A J ORNCZL NUHLG/CR tT87 VG2 TIMING ANALYSIS Of PWR FUf L PIN E

NUHrG/CH $672 ORNOIL A UNff E ttEMt NT ME SH GENf hATOR NI /C

• SK E A N Oil A (dW PRESSUHilLD IDH NOllt E-CYUNDL H INTi.RSLC110NS CONT AINING INNE H' WATFH RC ACTOR. EFF(CTS OF FIRE PROTECitON SYST(M AC-COHNE R CR ACKS TUAllON ON SArETY-HE LATED L QUlPMENT Evaluation Of Genenc Occupational E aposure Inue $7.

NUHEG/CH 4409 V04 DA1A BASL ON DOSE HE DUCT ON Rf. NUREG/CR 5819 PROOABftl1Y AND CONSEQUENCES OF RAPID LEARCH PROJECTS FOR NUCLE AH ROWER PL ANib. BORON DtLUTON IN A PWR A Scop *ng Stafy Operating Eapertence Periormance History NUHf G-1272 V06 NOI. Of FCL f OR ANALYSIS AND EVALUATON NUREG Or 1214 RIO HISTOHCAL DAT A buMMARY OF THE 9YSTE M OM 4ATIONAL DAT A 1991 Annual Npryt Powtd Nuctor6 IC ASSELSMENT Of LICENSCL PERFORMANCE.

NUHFG 1272 V06 NO2 Of fICE FOH ANALYSIS AND EVALUATION OF opt RATIONAL DAT A 1991 Annual Report Norveactors Petitione For ftulemaking NUHEG/CH.4ti19 V02. AGING AND St HVICE WEAR Or SOLf NO!D- NUREG0936 V11 N02. NHC REGULATO4Y AGLNDA.Quarterty Orf RAlfD VALVLS USED IN SAF EiY SYSTLMS OF NUCtrAR ReporLApril June 1992 POWL0 Pt AN1h Ivaluahon Of Morvtraig Methods RE ACTOR INSTRUMEN NUHf G/CR 5700 AGING ASSE SSMENT Of Phenomenology T AllON AND PROILC110N SYSTEM COMPONENTS Aging helated NUREG/C0 5305 V01: INTEGFLAl[O HISK ASSESSMENT FOR LA-Operating I mpetences SALLE UNii 2 HUCtf AR POWER PLANT Phenemmenotray And Risk NUHEG/CR 6/79 VQ1 AG8NG Or NON POWLR CYCLE HEAT Ex- Uncertamty Evalunton Pety,) ram (PRUtP)

CHANGERS USE D IN NUCL. EAR POWCH Pt ANTS Orcrating E pe*

ence Arvi F adure identdication Pipe Support Operahal Data NUREG/CH 54to TECHN! CAL EVALUATON OF GENEHlC ISSUE 113 NUREG/CP-0124 WORKLHOP 04 THE USE OF PRA METHODOLOGY DYNAMC QUAUDCATON AND TESTING OF LARGE BOHE HY-DRAUUC SNUDDE RS IOR THE ANALYSIS OF REAC109 EVENf 3 AND OPERAtlONAL DATA

@9 Operathonal E vent NUREG/CR-4599 V02 Nt: SHORT CRACKS IN PIPtNG AND PlPING MLM Senwannual Rem Apnt September 190 NUREG/CH 4674 VI$ PALCURSORS TO POTENTIAL SEVERE CORE DAMAGE ACC DENTS 1991 A ST ATUS HLPOHT Main Report And NURE G/CHD64 THEORE TICAL AND USER'S MANUAL FOR PC-FRAISE A Probatulistic Fracture Mechanics Computer Coda For Papeng 3 %4 NU G/CH-4674 V16. PRECURSORS TO POTLNTIAL SEVERE CORERaatelity Analysis DAMAGE ACCOENTS 1991 A ST ATUS RLPORT Apperukes B C.

And D. Pip 6ng System NUHEG/CR 5646. PnPING SYSTEM RESPONSE DUniNG HIGH LEVEL PC PRAISE SIMULATED SElSMIC TESTS AT THE HEISSDAMPFREAKTOR FA-NUHEG/CR 5664 THCORETICAL AND USER'S MANUAL FOR PCc CiUTY (SHAM TEST F ACallTY)

PRAISE A Probateshc I rectwo Mecharmcs Computer Code for Piping Hohatukty Analy9s. Plant Aging NUniG 1377 403 NRC RESEARCH PROGRAM ON PLANT AGING PRA LtSitNG AND SUMMARIES OF REPOHTS ISSUED THROUGH JULY NUREG/CP 0124 WORKSHOP ON THE USL OF PRA METHODOLOGY 1992 FOR THE ANALYSIS Or REACTOR LVENTS AND OPLRATONAL DATA. P6ug NUHLG/CH-4632 V01. ANALYSTS Of THE LASALLE UNt12 NUCLE AR NUREG'CR St#5 SEALING PERFORMANCF OF DEN 10 NITE AND POWER PLANT RlSK METHODS INTEGRAYiON AND EVALUATON BENTON11E/ CRUSHED ROCK BOREHOLE PLUGS.

PROGR AM (RMit P) Summary

l. 32 Subject Index b8 NUREG/CR4854 UNIVE RSAL . TRE ATMENT OF PLUMLS ANDNUREG/CR 6378 AGING DATA ANALYSIS AND RISK ASSESSMENT-STRESSES FOR PRESSURIZED THERMAL SHOCK EVALVATONS DEVELOPMENT AND DEuoNSTRA1 TON S10DY.

Pwtat Mmor NUREG/CR 6896 AUXILLARY FEEDWATER SYS1EM RISK-BASED IN SPECTION GUIDE FOR THE ST. LUCIE UNIT 1 NUCLEAR POWER NUREG/CR $868 DEVELOPMENT OF POslTON SENSITIVE PROPOR' GENERATON ST ATON TONAL COUNTERS FOR NUREG/CR 5905 REVIE W AND DEVELOPMENT OF COMMON NO.

AND PORTAL MON 110Rk, HOT PARTICLE DETECTON IN LAUNDRY MENCLATURE FOR NAMING AND LADELING SCHEMES FOR PHOBABILtSTIC RISK ASSESSMENT.

Power Osttilat6on f4UREG/CR 5910; LOSS OF ESSENTIAL SERVICE WATER IN LWRS NURE G/CR 6003. DEf4SITY WAVE INSTAblLITIES IN BOILING WATER(Gl.163) Scopng Study NEAUONI Probablinetic Safety A66*esment Pract6ce And Procedure Digest NUREG/CR 6667: APPROACHES FOR AGE DEPENDENT PROBABILt$.

NUREG4386 D00 R03 UNITED STATES f40 CLEAR FIEQULATORY TIC SAFETY AS$ESSMENTS WITH EMPHASIS ON PRORITIZATON COMMISSION STAFF PRACTICE AND PROCEDURE AND SENSITIVITY STUDIES D6GE ST commise6on. APpeat - Board And Licensmg Board DecitsonsJuly 1972 + Septenter 1991 0'8* " "*"*' P'#I Pressure Vessel NUREG!CR 5768 V02: FITNESS FOR DUTY IN THE NVCLE AR POWER i INDUSTRY Annual Summary 01 Program Performance Reports CY NURE G/CR 33PO V02- LWR PRESSURE YESSE L SURVEILLANCE [0- 1990 S1ME1RY tMPROVEMENT PROGRAM PSF Startup Esponrients NUREG/CR be59 Protection System MODELING 1HE IN4UENCE OF IRRADIAT ON TEMPERATURE AND DtSPLACEMLNT RATE ON RADIATION.IN. NUREG/CR 6700: AGING ASSESSMENT OF REACTOR INSTRUMEN-DOCED HARDLNING IN FERRMIC STEELS TATION AND PROTECTON SYSTEM COMPONENTS Agmg Related Preneurtaed thermal Shock NUREG/CR 5793 A COMPARISON OF ANALYSIS METHODOLOGIES Protective Moseure FOR PREDICTING CLE AVAGE ARREST OF A DEEP CRACK IN A RE, ACTOR PRESSURE VESSEL SUBJECTED TO PRESSURilED-THER. NUREG 1442 RO1: EMERMNCY RESPONSE RESOURCES GUIDE For Nuclear Power Plant Emergeneins.

MAL SHOCK LOADING CONDt10NS NUHEGICR 5854 UNfVLRSAL TREATMENT OF PLUME S AND Pump STRESSES FOR imESSURIZFD 1r(ERMAL SHOCK EVALUATIONS NUHEG/CF14886 EXPERIMENTAL At4D ANALYTICAL INVESilGATON NUREG/CP 0123 PROCEEDINGS OF THE SECOND NRC/ASME SYM-POSiUM ON oVMP AND VALVE TESTING He'd At The Hyatt Regency OF SELS THE SHALLOW-FLAW EFFECT IN REACTOR PRESSURE VES. Hotel,Wavangton DC, July 21-23,1992 Pressurised Water Reactor * ""'

NUHEGiCR 5787 V01- 11 MING ANALYS!S OF PWR FUEL PIN '

F ALLURE S Fmal Nport Mam Text And At ndees A J RELAP6/ MOD 2 I PWN IUEL F A LORf F l tA K NUREG/lA 0054 ASSESSMENT Or RELAP5/ MOD?, CYCLE 36.02, NUREG/CR $790- RISK EVALUATION FOR A B&W PRESSURilED USING NEPTUN REFLOODING EXPERIMENTAL DATA R A 088. POST TEST ANALYSIS AND NODALIZATION STUD-WATER REACTOR. EFFECTS OF FIRE PROTECitON SYSTEM AC.

flON ON SAF ETY.RELATED EOUlPMENLEvaluaton Of Genonc

'"" * '" "^"$

fY36 2 NUREG/CR 6819 PROBABILITY AND CONSEQUENCES OF RAPID Radioactive Maternal DORON DlLUTION IN A PWR.A Scopmg Stud /

NUREG/CR-2007 V10 RADIOACTIVE MATERIALS FIELE ASED FROM Price-Anderson Act NUCLEAR POWER Pt. ANTS Annual Report 1989. #

NUREG 1457; RESOURCES AVAILABLE FOR NUCLEAR POWER PLANT EMERGENCIES UNDER THE PRICE ANDERSON ACT AND "d*8"*I I"W l THE RODERT T. STAFFORD DISASTER RELIEF AND EMERGENCY NUREG/CR-6849 ORF FC: MANUAL FOR CONDUCTING RADIOLOGL 1 ASSISTANCE ACT. CAL- SURVEYS IN SUPPORT OF LICENSE TERMINATION. Draft l Report for Comment. I Probabilistic Risk Analyste NUREG/CP 0124. WORKSHOP ON THE USE OF PRA METHODOLOGY Reactw j FOR THE ANALYS:S OF REACTOR EVENTS AND OPERATONAL NUREG/CR4673 V02: TRAC-PF1/ MOD 2 CODE MANUALUser's Guide.  ;

DATA. NUREG/CR 5673 V03 TRAC PF1/ MOD 2 CODE MANUALProgeammer's Guede.

Probabilletic Risk Assessment NUREG/CR 4832 V01: ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR Neactor Containment POWER PLANT; RISK METHODS INTEGRATION' AND EVALUAllON NUREG/CR 5564= CORE CCNCRETE INTERACTONS USING MOLTEN FHOGRAM (RMiEP) Summary UO(2) WITH ZlRCONIUM ON A BASALTIC DASEMAT,The SURC 2 NUREG/CR-4832 V??" ANALYSIS OF THE LASALLE UNIT 2 NUCLEAR Espenment POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION IHOGRAM (RMIEP) Integrated Ovantification And Uncertainty Anaty. Reactor Cooitng System sis NUREG/CR4819. PROBADILITY AND CONSEQUENCES OF RAPID NUREG/CFb4832 V03 P1: ANALYSIS OF THE LASALLE UNIT 2 NV. BORON DILUTION IN A PWR.A Scoping Study -

CLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVAL-UATON PROGRAM (RMtEP) lniemal Events Acccent Sequence Reactor Event Quantifraton Man Repor.t NUREG/CP 0124. WORKSHOP ON THE USE OF PRA METHODOLOGY NUREG/CR4832 V03 P2 ANALYSIS OF THE LASALLE UNIT 2 NU. FOR THE ANALYSIS OF REACTOR EVENTS AND OPERATIONAL CLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVAL. DATA.

DATON PROGRAM (RMIEP) loternal Events - Aceglent Sequence OuantAcahon Appendices. Reactor instrumentation NUREGICR4832 V07; ANALYSIS OF THE LASALLE UN!T 2 NUCLEAR NUREG/CR 6700: AGING ASSESSMENT OF REACTOR INSTRUMEN-POWER PLANT. RISK METHODS INTEGRATION AND EVALUATON TATON AND PROTECTION SYSTEM COMPONENTS Agmg Related PROGRAM (GEP) External Event Sco . Quantificat ort Operateg Exponences.

NUREGICR4639 METHOOS FOR EX E NAL EVENT SCREENING QUANTIFICATION. RI.SK METHODS INTEGRATION AND EVALUA. Reactor Pressure Vesset {

TION PROGRAM (RMIEP) METHODS DEVELOPME NT, NUREGICR 5793, A COMPARISON OF ANALYSIS ME'THODOLOGIES  ;

NUREG/CR 5305 vot INTEGRATED RISK ASSESSMENT FOR LA- FOR PRE'DICTING CLE AVAGE ARREST OF A DEEP CRACK IN A RE.

SALLE UNIT 2 NUCLEAR POWER PLANT.Phenomenology And Rd ACTOR PRESSURE VESSEL SUBJECTED TO PRESSURIZED THER.

Uncertamty Evaluation Propam (PRUEP). MAL SHOCK LOADING CONDITONS.

l1 k

Subject index 33 NUHIG/CH %ti6 i APL10 MEN 1 AL AND At4ALYf 6 CAL INVE Sf 6 GALLON Safety System Of 1HL bHALt OW f ( AW f f f IC1 (N Hi AC10H PHI $$URI VL S NUf41G/CH 4b19 V02 AGING AND St HVICE Wf AH Of SOL .Notti Si ts OPl HAflD V ALVLS USE D IN SAF L TY SYSitMS C# NUGtf AR NUHi G/CH MO6 CONS T RAINT l FI l CI S ON I R ACT UHL 10UCM POWL H PLANT S [ vatuabon Of Msrutmng Mottoos f41 SS lOH CtHOUMI L HENilAltY OHlINf( D CHACKS IN HI AC-10H PfilSSURY VibSILS Screemng Procedure NUHIG/CH 48:12 V07 ANALYIWS Of THE L ASALLI UNIT 2 f40Cli AH Heactor Salety POWf R PLANT ALSK MF 1 HODS !NilGHAllON AND Lv AL UAtlON NUHL G/CH a4% Vo4 DATA DAS( ON DOSL Hi DUCllON HL pHOGp AM (HMit Pl L stwnal i vent Scolmog Quant 4 ation Sl AHCH PIOJECl1. f 04 Nt0Li AH POWL R PLAN 15 NUHf O/GR 4839 Mt 1 HODS iOH t ATE HNAL t VL N1 SCRLIN:NG OUANilF ICA160N RI5K Mf 1 HODS IN1( GHA160N AND L VALVA-H884tof V'at'l 110N PHOGH AM (HMit P) Mt 1 HODS Of vl LOPM[ N1.

NURE G/CH $667 GRADit NT STUDY 06 A L ANGL WE L D JOINING TWO IOIOtu A W6 SHL L L S Of THE MfDt AND HL AC104 Seatuig Performance VE S',I l - NUHf G/CH bfM M Alf4G PE Hf 00MANCE OF ULNTONilt AND NUHL U/CH Sf)91 Afit L( HAli D IHRADIAllON 1L S1 OF GUNDH(M DE N10NH t /CHUSHi O HOCK DOfM HOLI Pt UGS MWGI N HL AC10H VI SSL L THEPAN MAIL HlAL Seismic Hasard Recirculation NUIL v14tt: $1 AFI itCHNICAL POSillON ON INVf $1tGA1 ONS TO NUHLG/lA&wa HlCIHCUL A10N SUCllON LAHGL OHr AK LOCA IDE Nilr Y F AULT DISPL ACIMt NT HAIAHDS AND SFISM6C HAl-ANAL Yhts Of Dif. SANT A MAHIA DE GAHONA NWLt AH POWI H AHDs AT A Gt OLOGIC f tLPOSl10HY.

Pt ANI UNNG 1H AC lif 1(G1J1)

Severe Accident NI /lA4x)42 DISPifELD F LOW FILM DOtttNG An investi9aton Of W Alf R NUCL[ AH POWE A Pt ANTS Dratt Hefet F or Comment the Powt4ty to improve the Maiots ing4monh.d in 1ho NHC Com NUHC G /GP4120 PHOCl i DINGS 08~ THE IIf TH WOHPIJiOP DN poter Ehios Ior ihn Hofkuhng Phase Of It+ LOCA CONT AINMi N1 INTE GHITY. Hou in Washengton.DC May 12 14.1992 NUHE G/tA 0054 ASST hhMt NI Of f4LL APS/ MOD 2. CYCLE 36 02. NUHL G/Cil 240 F ASTGH ASS A ME CH ANISilC MODI t f 04 1HL USING NE PTUN HLf LOODING (3f'l FOME NT AL [P T A PHLD6CTON 05 XI,1. CS 10. DA. AND SH HlLL ASL F HOM NU-Cit AH IULL UNDlH NDHMAL AND SEVE RE ACCIDE NT negulatory Agenda COND@NS We Me f or Meinham Workstation, And Personal NUHLG 09 % V11 ff)2 NRG HL GUL A10H Y AGE NDA Ouartwly Compuun ApplicaNos Iktport. Aprel-Jurm 1992

$hallowflaw Effect Regulatory And Techn6 cal fieporg NUHEG/CH bf*6 LKPf RIMI Nf AL AND ANALYllCAL IN/f bilGAtlON Nuhl G 0304 V17 NOI. Hf GUl.ATOHY AND TECHNICAL Hl f DHis Of THf SHALLOW 4L AW (f f f Cf IN HC ACTOR PHf SSUH( Vt S.

(AHSTRACT INDiX JOURNAt) Compdatw>n i or i ust Quarte, M LS 1992. January March Shor1 Crack fleport To Congress NUH0G/CR-4%9 V02 N1 f> ORT CHACKS IN PlPING AND PIPING NUHf G lX)90 V1S Not H[POHT 10 CONGHE SS ON ADNOHMAL M W Smannual hgm he %ptww 1991 OCCt)Ritt NCE $ January March 1997 NUHL G OO90 V1$ NO2 HIPOHT 10 CONGHESS ON ADNOHMAL Simulated Seism 6C Test OCCUHHi NCI $ Apot. June 1992 NUHiG/CR %46 PIPING fiYST[M fit SPONSL DUHiNG HIGH (( VLL HoposHory Operat6onal Crttena SIMULATLD SLISMIC 1LSTS At THE HLl5SDAMPIHEAkf 0H F A-NUHfG/CH b604 RLPOSITOHY opt RAfiONAL. Chlfi klA Analysis Cit f TY (SHAM 1LSI F AClll1Y) filsleBased inspectiorn Outd, Software Standard NUHf G/CH Wun AVAttIAR) iI( DWATLR SYSTL M ThSK HAS( 0 IN, NUHLG/CR %30 HIGH IN1EGH1IY SOf TWARE ST ANDAHDG AND SPt CllON GUIDE FOH THE S t LUCIE UN11 1 NUCLI AH IUW[ R GUIDEllNC S GI NI HAllON S1 ATION Soleno6dOperated Valve fiock NUHEG/CR 4619 V02 AGING AND SC HVICL WL AH Of f 0LENOlO-NUHl.G/CH 5687. DOHIHOLE ST ADill1Y IN DiNstly WEtDED opt RAltD VALVES USED IN SArt TY SYSit MS OF NUCll AR POWC R PLAN f S C valuation Of Marutunng Methods TUFFS SolLi Weste Disposal Dules NUHCG 006 Vit NO2 NHC Hf GULATOHY AGE NDA Oaartedy NUHf G/CH 2907 V10 HADIOAClivt V ATERIALS HCLE ASCO FROM NUCLE AR POWL R Pt ANT S Annual Heport 1969 Heport,Apol June 1992 Hulen Of Practice Source 1erm NUHEG 0386 000 H03 UNNED ST ATI S NUCLE AR HLGOLATORY NUHLG-1465 DOf f FC ACCIDENT SOURCE TERMS FOR UGHT-AND Pf0CfDURL W ATL H NUCLE AH POWER PL ANT S Draft Neport f or Comment LOMuiSSION STAFr PR ACTICE DGLS1 Commmnion. Appeal Boom And L Kensing Board Ducrwons July 1972, Septemter 1991 StairWs4 Steel Standby Liquid Control System SHAM Test f acility NUDI G/CR M46 PIPING SYSILM RE SlON$E CVHING HIGH L[ytt NURL G/CA 6001 AGtNG ASS (SSMENT Of HWR ST ANDf3Y LIQUID SIMUL ATLD SEISMIC 1ESTS AT 1HL HriSSDAMPrHF AK10R FA- CONTROL SYSTEMS CluTY (SHAM TES1 FACIUTY) Startup E speriment SURC.1 NURLG/CR 3320 V02 LWH PHESSURE VESSI L SURVEILLANCE DO-SIME TRY IMPHOVEME NT PHOUR AM PSF Startup Esporrnents NURf G/CR 5443 COHE CONCHETE INTERACTIONS USING MOlTE N URANIA W11H llHCOfduM ON A tlMESTONE CONORLTE Stress Corroshon Crack 6ng DAM MAT The SURC-1 Espentnant NUREGICR 4667 V14 E NVlHONMLNT Atty ASSIS1ED CRACKING IN LIGHT W AT[R RE ACTORS Swmeannual Heport.Octatier 1991 Mrch Safeguarcia Summary Event Ust NURt G C625 V02 S AT EGU ARDS SUMMAHY EVfNT LIST 1992 (SSEL)Janusy 1.1990 Through Decomtww 31.1991 Substance Abuse Safeguards Surnmary Event Llal NUREG/CR SM8 V02 F11NLSS f OR DUTY IN THE NUCLE AR POWER INDUSTRY Annual Summary Of Prcxpam Performance Reports CY NUHEG OS25 V01 SAFf GUARDS SUMMAHY EVENT LIST (SSEU he=

440 Through Def.emtart 31,1989 1991

34 Subject Index Systernetic Assessment Of L6consee Performance NUR(GOS40 V14 N36 T11LE tiST OF DOCUMENTS MADE PUBLICLY NUh!G 1214 H10 HIS TORICAL DAT A SUVM ARY Of T HE SY S T EM A1 A V AIL A BL E . Jane 1 33 1992 IU A%ESSME NT OF L K t.NSEE Ff R8 OHM ANCE NUHEGOS40 V14 NN. TITLE tlST OF DOCUVENTS MADE PUBLICLY AVAILAllLE Jury 1 31,1932 NURE G OB37 V12 N92 NRC TLD DiR[CT RAD:ATON MONiTOH.NG Uncerta6nty Analy sis NE TWDHK Puyress Report Aped June 1932 NUPEG/CR 4tt32 V02 ANALYS!S Of THE LASALLE UNIT 2 NUCLE AR POWER PLANT RISK Mr THODS INTEGRA1 TON AND EV ALUATON TR AC-BF 1 pogggAy (ny EP) Integrated OvantAceiton And Uncertamty Anaty-NU4f GNA.0067 DECIRCULATON SUCTON LARGE DREAK LOCs n ANALYSIS OF THE SANT A MARIA Dl GARONA NUCL E AR POWE R PLANT US;NG TRAC-[410tJ1) Uttty Requirements Document NUMEG 1242 V01 NRC HEviEW OF ELECTRIC POWER RESE ARCH 1 R AC-BF l/ MOU 1 INSTITUTE'S ADVANCED LIGHT WATER RE ACTOR UTIUTY RE-NUMEG/CH43% V01 THAC BF 1/ MODI AN ADVANCE D BEST4 Sit- OUjREMENTS DOCUVf NT Pr MAf( COMPUT E R PHOGRAM FOR DAH ACC! DENT NURE G 1242 VD2 P01 N4C am SemabECTRIC VIEW Or POWER RE.

AN ALYSIS M@l Deschpton EE ARCH INSilTUTE S ADVANCED E GHT WATER RE ACTOR UllLl-NUAf.0/CH-43% V02 TRAC Of f / MOD 1 AN AD/ANCED Bf ST.CSTI TY HEOU REMENIS DOCUVENT. Evolutionary Plant Desgos Chapter MATE COMPUTER PROGRAM I OR BOfLING W ATER R( ACTOR AC. 3 CIDE NT AN ALYSIS Uw s Gsed' NUHEG 1242 V02 P02 NRC REVIEW OF ELECTRIC POWER RE-TRAC PF1/ MOD 2

" ^

NUnf G/UI 5F73 V02 TRAC PF1/ MOD 2 CODE MANUAL User's G#de N NUHLG/CH %73 V03 T H AC Pr 1/ MOD 2 CODE MANUAL Petgramrmn s Mrs Chasers 213 U# Valve TH AC/BF1 MOD, NUREG/CP4123 PROCFEDtNGS OF THE SECOND NHC/ASME SYM-NUHlG/CH 4391 TRAC /bF1 MOD 1 MODE LS AND CORREt ATIONS f*EIUM ON PUMP AND V ALVE TESTINO Held At The Hyatt Regency Hotel WasNngton.DC, July 21 23.1992 Thermal Sleeve NURl G/CR 6010 HIS TOH V AND CURRE NT ST ATUS OF GENERA. Vendor inspect 6pn fiON 3 THERM AL SLt E VES IN WL STINGHOUSE NUCLEAR POAER NUREG 0040 V16 N02 Lh'E NSEE CONTRACTOR AND VENDOR IN-PLANTS SPE CTION STATUS REPORT. Quaderly Report, Apnl June 1932.(Wheto Book)

Thermal Stress NUht G/CR $854 UNnil RSAL TREATMENT OF PLUME S AND Weld STRESSES F OR PHESSUH7ED THERMAL SHOCK f VAL UAf TONS NUHEG/CR 4599 V02 N1 SHORT CRACKS IN PIPING AND PIPING WE LDS Som>annuel Report Apol Septomta 1991 Thermal-Hyrirsubc NU4tG!CH 5867. GRAD:ENT STUD'r Or A LARGE WELD JOINING NURf G'CR fw9 V01 - DEVELOPING AND ASSESSING ACCIDENT TWO FORGID A 508 SHELLS OF THE MIDLAND REACTOR MANAGEMLNT PLANS FOR NUCLE AR POWE R VESSEL Pt ANT S Development Prxess And Cntena bi NUHL G/CH Re> V02 Ut VE L OPING AND ASSESS!NG ACCll'E N T Welded Tuff NUREGICR-5687: DORE HOLE ST ABILITY IN DE NSELY WELDED MANAGEMEN1 PLANS F OH NUCLE AH POWE R PLANTS Ev6iuaton CH A Protsype Pnx ns TUFFS Thermo!urninescent Dosamet r Tirconium NUR[G 0837 V12 NO2 NHC TLD DIRECT RADtATION MONtTOntNG NUREG/CR 6443 CORE CONCRETE INTERACTIONS USING MOLTEN NETWOhk Progrns Ropon Apol June 1992 UR AN'A Wi1H ltRCON'UM ON A LIMESTONE CONCREi E BASEM AT 1he SURC 1 E xpenment Tihe List NUHE G 'CR 5564 CORE-CONCRETE INTERACTIONS USING MOLTEN NUHE GC640 V14 NOS TillE LIST Or DOCUME NTS V ADE PUBLICLY UO(2) WITH ZlRCONIUM ON A t1ASALTIC BASEMAT.The SURC-2 AV AILADLE May 131,1992 g,,n nent.

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NRC Originating Organization index (Staff Reports)

This index lists those NRC organizations that have published staff reports. The index is ar-ranged alphabetically by mafor NRC organizations (e.g., program offices) and then by sub-sections of these (e.g., divis ons, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). Il further information is needed, refer to the rnain citation by NUREG number.

ADYlSORY COMMITTEE (S) NUREG 0525 V02. SAFEGUARDS

SUMMARY

EVENT LIST ADylSORY COMMITTEE ON NUCLE AR WASTE (SSEL) January 1.1990 Ttwough Decemter St.1991.

NUREG 1423 VD3. A COMPILATION OF REPORTS OF THE ADVISO- OlVISION OF HIGH-LEVEL W ASTE MANAGEMENT (POST 87D413)

AY COMMITTEE ON NUCLEAR WASTE.* dy i991 June 1992 NUREG-1451 STAFF TECHN! CAL POSITION ON INVESTOATIONS TO IDENTIFY FAULT DISPLACEMENT HAZARDS AND SEISMC I OFFICE OF EXECUTIVE DIFIECTOR FOR OPERATIONS (EDO) HAZARDS AT A GEOLOGIC REPOSITORY. l REGON 1 (POST 820201)

NUREGCM31 V12 NO2. NRC TLD DIRECT RADIATION MONITOR;NG U.S. NUCLE AR REGULATORY COMMISSION NETWORK Prmpens Report. Aphl-June 1992 DrrtCE OF THE GENERAL COUNSEL (POST B60701)

OFC OF ENFORCf MENT POST 8/0413) NVREG4386 D06 R03 UNITED STATES NUCLE AR REGULATORY NUREG4940 Vil NO2 ENFORCEMENT ACTIONS S GNiFICANT AC- COMMISSON STAFF PRACTICE AND PROCEDURE TONS RE SOLVED Ouarterty Progress Report.Aprdaune 1992. OIGEsT. Commission. Appeal Board And Ocensing Board Damons JW M72 SepWnW M EDO . OFFICE OF ADMINISTRATION (PRE 870413 & POST $90205) NRC + NO DETAlLED AF flUA00N GIVEN Of FICE OF ADMINISTRATION (POST 890205)

NUREG-1145 VOR U.S NUCLEAR REGULATORY COMMISSION NUREG 1442 RO1: EMERGENCY RCSPONSE RESOURCES 1991 ANNUAL REPORy GUIDE For Nuclear Power Plant Emergenciet DrVISON OF FREEDOM OF INFORMATION A PUBLICATIONS SERV. NUREG/CR-5416. TECHNICAL EVALUATION OF GENERIC ISSUE ICE S (POST 890205 113: DYNAMIC QUALIFCATON AN9 TESTING OF LARGE BORE NUREG-0304 V17 N01: REGUL.ATORY AND TECHNICAL REPORTS HYDRAllLIC SNUD8ERS (ABSTRACT INDEX JOURNAL) Compwition For Fast Quarter M92.JanuaTMarch EDO. OFFICE OF NUCLEAR REGULATORY RESEARCH (POST 820405)

OlVISON OF ENGINEEH:NG (POST 870413)

NUREG-0540 V14 N05 TITLE LIST OF DOCUMENTS MADE PUGLIC- NUREG 1377 803. NRU RESEARCH PROGRAM ON PLANT AGING.

NUR 0* LISTING AND SUMMARIES OF REPORTS ISSUED THROUGH 14 06 IT't L ST OF DOCUMENTS MADE PUBLIC-N EGO VE 7. ST OF DOCUMENTS MADE PUOUC-N C 122 Vot: PROCEEDINGS OF THE AGING RESEARCH g INF ORMATION CONFERENCE.

NUREG/CP0122 V02; PROCEEDINGS OF THE AGING RESEARCH NU EG '3 d CLEAR REGULATORY COMMISSON IS-N EC V5 REG b " ^

TORY COMMISSION IS-NtREG 36 1 2 NRG R GUL TORY AGENDA.Ouarterty WB AM DAu W W M W ME ReportAprOune 1992' MENT DEVELOPMENT AND DEMONSTHATION STUDY.

DIVISION OF SAFETY ISSUE RESOLUllON (POST 880717)

NUREG4933 S14 A PRIORIT!2ATON OF GENERIC SAFETY EDO. OFFICE OF THE CONTROLLER (PRE 820418 & POST 890205)

NU EG 147 V IN AL F C R S ANNUAL REPORT H EG 65 DAFT FC: ACCIDENT SOURCE TERMS FOR LIGHT.

1992.

W ATER NUCLExa POWER PLANTS. Draft Report For Comment.

EDO + OFFICE FOR ANALYSIS & EVALUATION OF OPERATIONAL EDO. OFFICE OF NUCLEAR REACTOR REGULATION (POST 800 DATA ASSOCIATE DIRECTOR FOR ADVANCED RLACTORS & LICENSE R OFF E OR ANALYSIS & EVALUATION OF OPERATIONAL DATA. DI-N 91 8 NUA REVIEW OF ELECTRIC POWER RESEARCH NUREG4090 V15 N01: REPORT TO CONGRESS ON ADNORMAL INSTITUTE'S ADVANCED UGHT WATER REACTOR UTILITY RE.

OUIREMENTS DOCUVENT Program Summary OCCURRENCES NURLG-0090 VIS NO2: January REPORTMarch TO 1992' CONGRESS ON ABNORMAL NUREG-1242 V02 P01: NRC RLVIEW OF ELECTRIC POWER RE-OCCURRENCES Apoldune 1992 SEARCH 1NSTITUTE% ADVANCED LIGHT WATER REACTOR NUREG 1272 V06 N01: OFFICE FOR ANALYSl$ AND EVALUATON UTluTY REQUIREMENTS DOCUMENT. Evolutionary Plant OF OPERATIONAL DATA 1991 Annual Report Power Reactors 0*8'gna Chapter 1 NUREG,1272 V06 NO2. OFFICE FOR ANALYSIS AND EVALUATON NUREG-1242 Vo2 P02. NRC REVIEW OF ELECTRIC POWER RE.

OF OPERATIONAL DATA.1991 Annual Report Nonreactors SEARCH INSTITUTE'S ADVANCED LIGHT WATER REACTOR DMSON OF OPERATIONAL ASSESSMENT (POST 870413) UTIUTY REQUtREMENTS DOCUMENT. Evolutionary P. ant NUPG1457: RESOURCES AVAILABLE FOR NUCLEAR POWER Designs Chapters 2-13 PLANT EMERGENCIES UNDER THE PRICE ANDERSON ACT AND DMSON OF SYSTEMS TECHNOLOGY (990827 921003)

THE ROBERT T. STAFFORD DISASTER REUEF AND EMERGEN. NUREG/CR 3950 V07: FUEL PERFORMANCE ANNUAL REPORT CY ASSISTANCE ACT. . FOR 1989. -

DIVISION OF SAFETY PROGRAMS (POST 870413) DIVISION OF REACTOR INSPECTION & SAFEGUARDS (670411 NUREG/CP-0M4 WORKSHOP ON THE USE OF PRA METHODOLO. 9210031 GY FOR THE ANALYSIS OF REACTOR EVENTS AND OPER. NUREG-0040 V16 NO2. LICENSEE CONTRACTOR AND VENDOR IN-ATIONAL DATA; SPECTION STATUS REPORT. Quarterly Report.Apnl + June 1992 (Whte Book)

EDO OFFICE OF NUCLEAR WATERIAL SAFETY & SAFEGUARDS DIVISON OF LICENSEE PERFORMANCE & QUAUTY EVALUATON OlVISON OF SAFEGUARDS & TRANSPOR f ATON (POST 870413) (8704 t t 921003)

NUREG4525 Vol. SAFEGUARDS

SUMMARY

EVENT UST NUREG 1214 Rio HISTORICAL DATA

SUMMARY

OF THE SYSTEM-(SSEL) Pre-NRC Through Decerrt>er 31,1969. ATIC ASSESSMENT OF LICENSEE PERFORMANCE 35

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NRC Originating Organization Index (International Agreements)

This index lists those NRC organizations that have published international agreement re-ports. The index is arranged alphabetically by major NRC organizations (e.g., program of-fices) and then by subsections of these (e.g., divisions, branches) where appropriate. Each entry is followed by a NUREG number and title of the report (s). If further information is needed, refer to the main citation by NUREG number.

LDO . OFFICE OF NUCLE AR REOULATORY RESE ARCH (POST 820405) NUREG/lACO67; RECIRCULATION SUCitON LARGE BREAK LOCA OFFICE OF NUCLE AR HEGULATORY RESEARCH (POST 8607PO) ANALYSIS OF THE SANTA MARIA DE GARONA NUCLEAR

>"' o IA4042. DISPERSED FLOW FILM DOILING.An investgatx>n POWER PLANT USING TRAC.BFitG1J1) w Posuuhty To improve The Models implemented in The NRC NUREG/tA 0088 POST TEST-ANALYSIS AND NODAU2ATION Computer Codes For The Refkgeng Phase Of The LOCA. STUDIES OF OECD LOFT EXPERIMENT LP42-6 WITH RELAP5/

NUREG/tA 00$4 ASSESSMENT OF RELAP5/ MOD 2. CYCLE 36 02. MOD 2 CY36-02 USING NLPTUN REFLOODING IXPEHIMENTAL DATA.

37

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NRC Contract Sponsor Index (Contractor Reports)

This index lists the NRC organizations that sponsored tho contractor reports listed in this cornpitation. It is arranged alphabetically by major NRC organization (e.g., program offico) and then by subsections of thoso (e.g., divisions) where appropriato. The sponsor organiza-tion is followed by the NUREG/CR number and title of the report (s) proaared by that organi-zation. Il further information is nooded, refer to the main citation by the NUREG/CR number.

EDO OFFICE FOR ANALYSl8 & EVALUATION OF OPERATIONAL NUR[G/CH SS64 THEORETICAL AND USER'S MANUAL FOR PC-DATA PRA!SE A Probatxhats f racture Mechancs Contvief Code for DIVISION OF SAf f TY PHOGRAMS (PO*if 670413) Piping Rehability Anava a NUHEG/CH 4074 V16 PHICUHSONS 10 POf ENTIAL SEVEHE NUHEG/CH bo6L GHADILNT STUDY OF A LANGE WELD JO'NING CORE DAMAGE ACCIDt NTS 19G1 A STATUS HEFORT Main TWO ICRG[ D A $08 SHELLS OF THE MIDLAND HEACTOR Ra$et And Apptends A yr SS[ L NUHEG/CH 4614 V10 PHECUH90HS TO POTEN11AL SEVERE NUREG/CR $8'? OHNOlL A fiNf f E ELEM(NT MESH CENT RA-COHE DAMAGE ACCOE NT S 1991 A STATUS TOR FOR NOZZLE 4YLINDE R INTLRSECTIONS CONTAINiNG HEPORT APsendeces B.C. And D INNE HCCANER CHACKS NUHf G/Cf1-bn86 EXPERIMLNTAL AND ANALYTICAL INVESTIGA-EDO. OFFICE OF INFORMATION RESOURCES MANAGEMEN1 & ARM TON OF THE SHALLOW FLAW E FFECT IN HEACTOH PHE SSURE (POST 881109) VE SS[ LS OrrlCE OF INFORMATION f tt SOURCES MANAGit ENT (POST NUHE G/Di $891: ACCEL.t HAT [0 IHRADIATON TEST OF OUN-

1. ' b "

J 1I 4P t A al R r 19 9 CONTHOL SYSTEMS.

NURtG/CH 6008 CONSTRAINT EFFECTS ON FRACTURE TOUGH-EDO . Of flCE OF NUCLE AR MATIHlAL SAF ETY & SAFEGUARDS OlVISON OF HIGH LE VEL WASit MANAGLMENT (POST B70413) NESS FOH CIRCUMFlHENTIALLY ORENTED CRACKS IN RE AC4 NUH[ G/CR4804 HEPOSMORY ODERATIONAL ChiTERIA ANALY- TOR PHESSURE VESSELS.

Sls DIVISION OF RE GULATORY APPLICATIONS (POST 8704131 NUREG/CH 4409 V04 DATA DASE ON DOSE HEDUCTION HE-EDO OF FICE OF NUCLEAR REQUL ATORY RESEARCH (POST 820405) SE ARCH PROJECT S FOH NUCL E AH POWE R PLANTS DIVISION OF E NGIN[ L HING (POST 870413) NUHLG/CH $685: SEALING PERf ORMANCE OF DENTONITE AND NUHEG/CH 3320 V02. LWR PHESSURE VESSEL SURVEILLANCE BE NTONITE/CHUSHE D HCCK DOREIOLE PtLICS DOSIMETRY IMPHOVE ME NT PROGRAM PSF Startup Expenments NUREG/CR $687, DOHLHOLE ST ADILITY IN DENSELY WELDED i NURLG/ul-4469 V13 NONDESThuCTfvE ERAMINATION (NDE) H[- TUFF S l LIADILITY FOR INSERVHZ INSPECTION OF LIGHT WAiER NUHEG!CH $849 DHF FC. MANUAL FOR CONDUCTING HADOLOG-l Hf ACTORS Som.aanual Report October 1990 March 100l- ICAL faVRVEYS IN SUPPORT OF UCENSE TERMINATION Droft NUHEG/CH 4469 V14 NONDLS)StVCTIVE EXAMINATION (NDC) RE~ Haport i or Comment LIAHrtlTY FOR INSERVICE INSPECTON OF UGHT W ATEH NUREG/CH $868 DEVELOPMENT OF POSITON SENSITIVE PRO-HE ACTORS Samtennuel neport Apd 1991 Septomter 1901 POPTIONAL COUNTERS FOR HOT PARTICLE DETECTION IN NUALG/CH 4599 V02 N1 SHORT CHACKS IN PtPING AND PIPING LAUNDHY AND PORTAL MONITORS.

N t /CF 4 4 ENV R N LLYA ISTED CHACKING g[ph)M T P IN LIGHT WATE R REACTORS Semiannual Report,0ctot>er 1991 ' DIVISION OF 6AF ETY ISSUE RESOLUTION (POST 880717)

M W92 NUREG/CR 4832 V01; ANALYSIS OF THE LASALLE UNIT 2 NUCLE-NUHE G/CH 4744 VO6 NI. LONG-TE AM EMBHIT~o.EME NT OF CAST AR POWER PLANT! Al5K METHODS INTEGRATON AND EVALUA-DUPtE X STAINLCSS STEELS IN LWR SYSTLMS Semiantual flON PROGRAM (RMIEP) Summary NUNEG/CH 4832 V02. ANALYSIS OF THE L/* ALLE UNIT 2 NUCLE.

NU t /C 48 9 AIkG O SL AVICE WE AR OF SOLENOID- ^" " " "

OPERATED VALVLS USED IN SAF ETY SYSTEMS OF NUCLE AR T ON PROGRAM (RMIEP) Infograted QuanlQaten And Uncertainty POWU1 PLANTS Evaluation Of Moniton Methods NUR[ G/CR-5378 AGING DATA ANAL IS AND RISK ASSESS- ^^"Y"

NUREG/CR 4B32 V03 PI: ANALYSIS OF THE LASALLE UNIT 2 NU-MENT-DEVELOPMENT AND DEMONSTRATION STUDY NUHEG/CR $$87. APPROACHES FOR AGE DEPENDENT'PROBABp CLEAR POWER PLANT: HISK METHODS INTEGHATION AND LISTIC SAFETY ASSESSMENTS WITH EMPHASIS ON PRIOntTilA, EVALUATION PROGRAM (HMiEP)lntertwl Events Accident S6-Quence Quantifmatm Mam Report TION AND SENSITivlTY STUDIES NUREG/CR 5646: PIPING SYSTEM RESPONSE DURiNG HIGH. NUREGICH-483I V03 P2. ANALYSIS Of THE LASALLE UNIT 2 NU-LEVEL SIMULATED SflSMiC TESTS AT THE HEISSDAMPFREAK, CLE AR POWER PLANT: R:SK METHODS INTEGRATION AND TOR FACILITY (SHAM TEST FACIUTY)

EVALUATON PROGRAM (RMtEP)lniemal Events Accident Se-NUREG/CR 5700- AGING ASSLSSMENT OF REACTOR INSTRU. quence Ouantifcation Appendeen MLNTATION AND PROTECTION SYSTEM COMPONENTS Agmg. NUHEG/CH 4M2 VOT: ANAL.YSIS OF THE LASALLE UNIT 2 NUCLE-AR POWER PLANT: RISK METHODS INTEGRATION AND EVALUA<

Related Operaung Dpanences TON PROGRAM (RMIEP) Enternal Event Scoping Quantifcatert NURLG/CR 5772 VUt- AGING, CONDITON MONITOR!NGi AND LOSS-OF COOLANT ACCIDENT (LOCA) TESTS OF CLASS 1E NUHEG/CH-4839 METHODS FOR EXTERNAL Evt NT SCREE NING Et f CTRICAL CABL ES Croanhnhuf Potyolefm Cat >ies OUANTIFICATION RISK METh3DS INTEGRATON AND EVALUA-NUREG/CR 57/9 V01- AGING OF NON POWER-CYCLE HE AT Ex. TON PROGRAM (RMIEP) METHODS DEVELOPMENT.

CHANGERS USED IN NUCLEAR POWEFt PLANTS Operatog Empo. NUREG/CR 5305 V01. INTEGRATED HISK ASSESSMENT FOR t A-renc-o And Fadure idenhficanon. SALLE UNIT 2 NUCLEAR POWER PLANT.Phenomennlogy And Risli NUREG/CH-5793 A COMPARISON OF ANALYSIS METHODOLO- Uncertainty Evaluation Proptam (PRUEP)

GiES FOR PREDICTING CLEAVAGE ARREST OF A DEEP CRACK NUREG/CR-5416 TECHNK,AL- EVALUATION OF GENERIC ISSUE IN A REACTOH PRESSURE VESSEL SUDJECTED TO PRESSUR. 113 DYNAMIC OUALIFICATION AND TESTING OF LARGE DORE IlED-THERM AL SHOCK L OADING CONDITONS HYDRAUUC SNUBBERS.

NUREG/CR 5810 EVALUATION OF MHTGR FUEL REllABILITY. NUPEG/CH 5767 V01: TIMING ANALYSIS OF PWR FUEL PtN NUREG/CH4859 MODELING THE INFLUENCE OF IRRADIATION F AiLURES Final Report Man Test And Apperoces A J TEMPERATURE AND DISPLACEMENT RATE ON RADIATION IN- NUHEG/CR 5787 V02. TIMENG ANALYSIS OF PWR FUCL PIN DUCED HARDENING IN FERRITIC STEELS. FAILUHES Foal Report Appendices K-L 39

- - - , _a . .

40 NRC Contrat

• Sponsor index NUREG/CR 5790 FilSK EVALUAT104 FEW A B&W PRESSURIZFD CtLAR FUEL UNDE R NORMAL AND SEVE RE-ACCIDENT WATER REACTOR. ElFECTS OF FIRE PROTECTION SYSTEM AC- CONDnIONS Users Gwde For Matt amo. Workstation. And Pe4on-TUATION ON SAFETY RELATED EQUIPMENT Evaluston Of Gener- al Comruter Appk at.ons e gave 57. NURLG/CA LB54 UNIVERSAL T RE ATMENT OF PLUME S AND NUREG/CR-5905 REVIEW AND DEVELOPMENT OF COMMON NO. STRE SSES FOR PRESSURl2ED THERMAL SHOCK EVALUA-MENCLATURE FOR NAMING AND LABELING SCHEMES FOR T 6ONS PRODABILISTIC RISM ASSESSMENT NUREG/CR 6931 HIGH INTEGRITY SOFTWARE STANDARDS AND l NUREG/CR-5910. LOSS OF ESSENTIAL SERVICE WATER IN LWRS GU'DE LINE S (Gi-153) Scornng Study NUREGICR4009 V01: DEVELOPING AND ASt :S$1NG ACCIDENT NUREG/CR-6010. HISTORY AND CURRENT STATUS OF GENERA. MANAGEMENT PLANS FOR NUCLEAR POWER TION 3 THERMAL SLEEVES IN WEST 6NGHOUSE NUCLEAq PLANTS Developreent Frocess And Cattes POWER PLANTS NUREG/CR 6000 V02 DEVELOPING AND ASSESSING ACCIDENT DIVISION OF SYSTEMS RESEARCH (POST 800717) MAN AGE MENT PLANS FOR NUCLEAR POWER NUREG/C4 4356 V01: TRAC DF1/MODl; AN ADVANCED DEST Es. PLANTS Evelvabon Of A Prototype Frocess l TIMATE COMPUTE R PROGRAM FOR BWR ACCIDE NT EDO. OFFICE OF NUCLEAR REACTOR REGULATIO. POST 000428)

Sv NU E C 3% 0 TRA Ft/ MOD 1.AN ADVANCED BESTf Sil. h0N[n, a9 y E9p(O g g C AN UAL REPORT MATE COMPUTER PROGRAM FOR BOtLING WATER REACTOR pop gggg ACCIDENT ANALYSIS User's GMe- NUREG/GR 6819 PROBABluTY AND CONSEONENCES OF RAPID NURE G/CR.4391; TRAC /BF1-MOD 1 MODELS AND CORRELATIONS NVALG/CR 5443 BORON utUTION IN A PWR A Sc<png Study CORE CONCRETE INTERACTIONS USING NDHE G/CR 6003 DENSITY WAVE INST ABILITIE S IN BOIL NG MOLTEN URANIA WITH ZlRCONIUM ON A LIMESTONE CON' WATER REACTORS CRETE DASEMAT.The SURGI Extetsment DIVISION OF REACTOR INSPECTION & SAFLGUARDS (870411+

NUREGICR 5%4 COnECONCRE1E INTERACTIONS USING 921003 MOLTEN UO(2) WITH ZlRCONIUM ON A BASALTIC DASEMAT,The NUREG/v)R4758 V02. FITNESS FOR DUTY IN THE NUCLEAR SURGP E mpenment POWER INDUSTRY Annual Summary Of Program Performarre NURE G/CR-5673 V02 TRAC PD/ MOD 2 CODE MANUAE. User t Rep. orts CY 1991 l Gunto DIVISION OF RADIATION PROTECTION & EMERGENCY PREPARED.

NUREG/CR %73 V03 TRAC. Pit / MOD 2 CODE NE SS (87041192100 MANUAL Programmer's Guide NUREG/CR 5896 AVAll6ARY FEEDWATER SYSTEM RISK BASED NUREG/CR-5640 FASTGRASS A MECHANISTIC MODEL F OR THE INSPECT *ON GUIDE FOR THE ST, LUCIE UN!T 1 NUCtEAR PREDICTION OF XE, I, CS, TE, BA, AND SR RELEASE FROM NU- POWER GENERATION $T ATION l

r 8

J

Contractor Index This index lists, in alphabotical ordor, the contractors that prepared the NUREG/CR reports listed in this compilation. Listed below each contractor are the NUREG/CR numbers and

titles of their reports if further information is noodod, refer to the main ci'ation by the NUREG/CR number.

i AMGONNE NATIONAL LABORATORY DNV TECHNICA NUREG/CR Af47 V14 ENylRONMENT ALLY ASSISTED CR ACKING IN NUREG/CH-tcl78 AGING DAT A ANAltSIS AND RISK ASSES $ MENT-UGHT WATi n RE ACTORS Semiannuai neportOctotier 1991 Ma'th DEV6 LOPMENT AND DEMONSTRAllON STUDY.

1992 NUREG/CH 4744 V06 N1. LONG-TERM EMDRITTLEMENT OF CAST [G& o lDAHO, lNC.

DUPL Ex S t AINLE SS STEftS IN LWR SYST( MS Semennual NURE G/CP-0173. PROCEEDINGS OF THE SECOND NRC/ASME SYM-Repor10cInter 1990 March 1991 POslVM ON PUMP AND VALVE TESTING Held Al The Hyatt Reponey NUHLG/GRM40 F ASTGRASS A MECHANISTIC MODEL F04 THE HdeLWuhir PREDICTION OF AE. L CS, TE BA, AND SR Rf LEASE FROM NU- NURE G/CR 4dy lonIO. M 2123.19926 V01" TRAC-DF1/ MOD 1: AN ADVANCE1 CLEAR FULL UNDER NOHMAL AND SEVERE ACCIDf NT MATE COMPUTER DROGRAM FOR BWR ACCIDENT -

CONDITONS User's Guxle For MLreme, WorkstatwA Arv$ Perenal ANALYSIS Model Deseviption Computer Apphcotona NUREQ/CR-4356 V02 TRAC DF1/ MOD 1.AN ADVANCED DEST-ESil-MATE COMPT!TER PROGRAM FOR DOlllNG WATER REACTOR AC-art 2ON A. UNIV, OF, TUCSON, AZ CIDE NT ANALYSi$ User's Gude NUREG/CR $685 SEAUNG PrRFORMANCE OF BENTONITE AND NUREG/CR4391: TRAC /BF1 MOD 1 MODELS AND CORRELATONS DENTONiit /CROSHf D ROCK HORFHOLE PLUGS NUREG/CR-5378 AGING DAT A ANALYSIS AND FUSK ASSESSMENT ~

NUREC/GR SM7. DORE HOLE ST ABluif IN DE NSELY WELDED DEVELOPMENT AND DEMONSTRATION STUDY.

T V" 5 NUREG/CR 5416 TECHNICAL EVALUATION OF GENERIC ISSUE 113 NUP.EG/CR $819 PROBABILITY AND CONSEQUE NCES OF RAPID DYNAMIC OUAUFICATON AND TESTING OF LARGE DORE HY.

DORON Dit UTION IN A PWR A Scoperv4 Stude DREC SNUDDERS NUDt O/CR-tB80. NONISOTHERMAL HTDROLOGIC TRANSPORT Ex. NUREG/CR 5646 PIPING SYSTEM RESPONSE DURING HIGH. LEVEL PERIMENTAL PLAN SIMULATED SEISMIC TESTS AT THE HEISSDAMPFREAKTOR Fo DARTHOLD & ASSOCIATES,INC. CILITY (SHAM TEST FACILITY [

NUREG/CR $810- EVALUAflON OF MHTGA FUEL RELIAblLITY NUREG/CR 5787 vot: TIMING ANALYSIS OF PWR FULL PIN F ALLURES Foal Report Main 1ert And Appereces AJ DAT1ELLE HUuAN AFFAIRS RESEARCH CENTERS NUREG/CR5787 V02: TIMING AN ALY SIS OF PWR FUEL PIN NUREG/CH $758 V02 F11 NESS FOR DUTY IN THE NUCLLAR POWER FAILURES Fmal Report. Appendices KL INDUSTRY Annual Summary Of Program Performance ReportsCy NUREQ/CR$905 REVIEW AND DEVELOPMENT OF COMMON NO.

1991. MENCLATURE FOR NAMING AND LABEUNG SCHEMES FOR PRODABluSTIC RISK ASSESSMENT, BATTELLE MEMORIAL INSilTUTE COLUMBUS LABORATORIES NUREQ/CR 6009 V01: DEV! LOPING AND ASSES $ LNG ACCOENT NUREG/CR-4599 V02 N1 SHORT CRACKS IN PIPING AND PIPING MANAGEMENT PLANS FOR NUCLEAR POWER WELDS Senwnnua! Report Apnt-Soptemter 1991, PLANTS Development Process And Critena NUREQ/CR4009 VOL DEVELOPING AND ASSESSING ACCIDENT BATTELLE MEMO *tAL INSTITUTt; PACIFIC NORTHWEST MANAGEMENT PLANS FOR NUCLEAR PCWER PLANTS Evaluaten LADORATORY Of A Prototype Process.

NUREG/CH 3320 V02: LWR PRESSURE VESSEL SURVEILLANCE DO-SIMETRY IMPROVE MEfC. PROGRAM PSF Startup Emperernents [NERGY, DEPT. OF NUREG/CR 3950 V07; FUEL PERFORMANCE ANNilAL REPORT FOR NUREG/CR-4391. TRAC /DF1. MODI MODELS AND CORRELATIONS 19A9

- NUREG/CR4469 V13. NONDESTRUCTIVE EXAMINATION (NDE) FIEll- FAILURE ANALYS68 ASSOCIATES,INC.

ABILITY FOR INSERVICE INSPECTON OF UGHT WATER NUREG/CR 5864 THEORETICAL AND USER S MANUAL FOi, PC ,

RE ACTOFIS Semiannual Report Octoter 1990. March 1991 PRAISE A Prottatahstic Fruture Mechanics Computer Code For Piping NUREGrCR 4409 V14 NONOESTRUCTIVE EKAMINATION (NDE) RElle n e ig An, %

ABILITY FOR INSERVICE INSPECTON OF LIGHT WATER RE ACTORS Semaannual Report April 1991.Septemter 1991. FEDERAL EMERGENCY MANAGEMENT AGENCY NUREG/CR !158 V02. $ 1TNESS FUFt DUTY IN THE NUCLE AR POWER NUREG-1442 R01: EMERGENCY RESPONSE RESOURCES GUIDE For INDUSTRY. Annual Summary Of Program Performance ReportsCY Nw.lomr Peer Plant Emergencsoa.

1991.

NUREG/CR4806 AUyluARY FEEDWATER SY!TFM RISK BASED IN' FINNISH CENTRE FOR RADIATION 4 h0 CLEAR SAFETY (FINLAND)

, SPECTION GUIDE FOR THE ST. LUCIE UNIT 1 NUCLEAR POWER NUREG/CR 5819. PROBABlUTY AND CONSEQUENCES OF RAPID NFE / A 6( 1 AGI ASSESSMENT OF DWR STAND 0Y LOUiD CONTROL SYSTEMS- HALUDURTON NOS ENVIRONMENTAL CORP.

• NUREG/CR 5787 Vot: TIMING ANALYSIS OF PWR FUEL PIN DROOKHAVEN NATION AL LABOR AT04 t. F ALLURES Foal Repor1 Mam Test And Appendices AJ

, NUREG/CR 2907 VI(r RADCACTIVE MATERIALS RELEASED FROM NUREG/CR 5787 V02. TIMING ANALYSIS W PWR FUEL PIN FAILURES Final Neport Apperdcas KC N / 4 VO A DA SE REDUCTON RE.

C! ARCH PROJEC19 FOR NUCLE AR POWER PL ANTS KANSAS UNIV OF' LAWNENCE KS f NURE G/CR 5819 PROBADILITY AND CONSEQUENCES OF HAPID BORON DILUTON IN A PWR.A Semng Stu@ NUREd/ cad 888 EXPERIMENTAL AND ANALYTICAL INVESTIGATION OF THE SHALLOW FLAW EFFECT IN REACTOR PRESSURE VES-CALIFORNIA. UNIV. OF, SANT A DARDARA, CA SCLS-4 NURE G/CR4854: UNIVE RSAL TREATMENT OF PLUMES AND STRESSES FOR PRESSURIZED THERMAL SHOCK LVALUATONS. El Cd O t UHrG CR 5443 CORE CONCRETE INTEGACTIONS USINC MOLTEN CENTER FOR NUCLEAR WASTE RE00LATDRY ANALYSES URANIA WlTH ZlRCONIUM ON A LIME STONE CONCFIF TE NU9EG/CR $804 REPOSITORY OPERATIONAL CRITERIA ANALYSIS. BASEMAT.The SURC 1 Esperwnent.

41

42 CoruraClor index t4URL~G/CR 554 CORE.CONCHETE INTERAC10NS U9NG MOLTEN S. COHEN & ASSOCIATES,INC, UCx2) Wif ri llFAQiUM ON A ilASALTIC BA$lMAT 1he SURC-2 NUM G/CR4010 HISTORY AND CURRENT ST ATUS OF GE NERA-E rpenment TON 3 THERMAL SLEEVES IN WESTINGHOUSE NUCLE AR POWER LAWRE NCE LIVE % 'OU NA16GUA. LADOHATORY NURE G/Cf' M64 THE ORETCAL AND USU S MANUAL FOR (C SANDIA NATIONAL LABORATORIES

'VtAiSL A Probatal@c Da9 fe R d.noxa Conwter Code For Pipmg NURE G/CP 0120 PROCEEDINGS OF THE FIFTH WORA$ HOP ON Rew. ley AnsW CONT AINME NT INTEGRITY. Held in Wodu 6n tC.May 12 14.1992 NUHEGICP-0124 WORkSHOD ON THE USE gU PHA ME THODOLOGY D# !.LAMOS NATIO%.L LADOPATORY FOR THE ANALYSIS OF REACTOR EVENTS AND OPERATIONAL NUtd G TA 5673 V02 TRAC P81/ LOD 2 CCSt MANUAL user s Gxte DATA PJURE GX kW3 V031 RAG PFbMC2 (COE MANUAL pc yfvnw s NUR[G/C"4432 V01 ANALYS!S OF THE LASALLE UNIT 2 NUCLE AR Ga4 POWER JLANT RtSK METHODS INTEGRATION AND EVALUATON "FOGR AM (DMif P) Summaq MARTIN tun 7JLtlit'O kRvicEL WC. NURL G/CR 4632 V02 ANALYS:S OF THE LASALLE UNIT 2 NUCLI AR NURE G/CR4010 H5f 0RV AND CURRENT STATUS Or GENERA- POWER PLANT. FilSK METHODS INTEGRATON AND EVALUATON flON 3 THERMAL SLt EVE S IN WEST.NGHOUSE NUCLI AR POA[.R FHOGK.t'd (RMiEP) Integated Quantifcation And Uncertainty Anary.

PU NT S em NUREG/CFi4832 V03 P1 ANALYSIS OF THE LASALLE UNIT 2 NU-M ARYLA ND, UNIV. OF, COLLE GE P ARK, MD CLEAR POWE R PLANT: l'SK METHOvS INTEGRATON AND EVAL-NURt G/CR 'M7 GRADIENT STUDY OF A LARGE WELD JOINWG UAl TON PROGRA M (RMiEP) intenal E vents Aatdent Sequence __

TWO FORGED A 506 SHELLS OF THE MOLAND REAC1DR Cannikaten Vain Rep.ott VE SSE L NUHLF./CR-4832 V03 F2. ANALYSl5 OF THF LASALLE UNIT 2 NU-CLE AR POWER PLANT: RISK METHODS INTEGRATION AND EVAL.

MATTRIALS ENGINEERING ASSOCIATES,1NC. VATON PROGRAM (RMIEP) lr tornal Events Aratont Sequence NUPt G/CR %91 ACCELE RATE D IRRADIAllON TEST OF GUNDhEM- Omntkahon Appon$ces MEN REACTOR V( $$f L 1REPAN MATERIAL NUREG!LR46M VU7. ANALYSIS OF THE LASAttE UNIT 2 NUCLEAR POWER PLANT: RISK METHODS INTEGRATION AND EVALUATION N ATtONAL INSTITUTE 08' ST AND AF.DS & TECHNOLOGY (FORMERLY PROGRAM (RMtEP) Enieens! Event Scopmp Ouantshcahon NATION AL BUREAU OF NURLG/CR 46W ME1 HODS FOR EX1ErtNAL EVENT SCREENING NURLG/CR SMO H.GH INEGRITY SOFTWARE ST ANDARDS AND OUANTIFICAllON R!SA METHODS INTEGRATON AND EVALUA.

GUIDELINE S TION PROGRAM (RMIE P) METHODS DE VELOPMENT.

NUREG/CR-5305 V01- INTEGRATED RISK ASSESSMENT FOR LA-NTS/SM A. INC. SALLE UNfi 2 NUCLEAR DOWER PLANT Phenomenology And Ribh NUMEG/ CRAB 32 V07. ANAL.YSi$ OF THE LASALL E UNIT 2 NUCLLAR Unctw+swity Evaluaton Pgfymm (PRUE P)

POWER P'_ ANT. R!SK METHODS INTEGRATION AND EVALUATON NUREG/CR 5443 COREw4CALTE INTERACTONS USING MOLTEN PROGRAM (HMIEP) f rienwl Event Scoping Ouantrhcaton URANIA WITH ZlRCONIUM ON A LIMESTONE CONCRETE NUREG /CR4839 METHODS LOR EXTE RNAL EVENT SCREENING BASE MAT.The SURC 1 Empenment.

QUANTIFICA TION RIM ME THCOS INTEGRATK)N AND F VALUA. NUREG/CR 5564 CORE CONCRETE INTERACTONS USING MOL1E N TlON PRCGRAM (RM1E P&ETHODS DEVELOPMENT, UO(2) WITH ZlRCONtVM ON A DASALTIG BASEMAT.The $URC-2 E openment OAK ROGE ASSOCIATED INIVERSITIES NURLG/GR-5772 V01' AGING. CONDITON MONITORING, AND LOSS.

NUREG/CR4649 DAF FC. MANUAL FOR CONDUCilNG RADOLOGl. OF-COOLANT ACCOENT (LOCA) TESTS OF CLASS 1E ELECTRICAL CAL SURVEYS IN SUPPORT OF LICE NSE T L RMIN ATION Draft CABLES Crnnsknkm1 Polyolehn Catdes Repo,t Foi Commed NUREG/CR 5790 RISK EVALVAllON FOR A B&W PRESSURIZED WA1ER REACTOR. EUECTS OF FIRE PROTECTION SYSTEM AC-OAK RIDGE NATIONAL EADORATORY TUAflON ON SAFETY RELATED COUIPMENT Evaluation Of Genene hvREG/CR 4674 V15 PRECURSORS TO POTENTIAL SEVE RE CORE issue 57 DAMAGE ACCIDENTS 1991 A STATUS REPORT. Main Report A%d NUREG/CR 5910 LOSS OF ESSENTIAL SERVICE WATER IN LWRS Appendra A (Gbi53)Scopeng Study.

NUREG/CR4674 V16 PRECURSORS TO POTENilAL SEVERE CORE -

DAMAGE ACCIDENTS 1991 A STATUS REPORT Appendices D*C' SCIENCE & ENGINEERING ASSOctATES lNC.

And D NUREG/CR 575r0 RISK EVALUATON FOR A B&W PRESSURilED WATER REACTOR. EFFECTS OF FIRE PROTECTION SYSTEM AC-NURE G/CR-4819 V02: AGING AND SERVCE WEAR OF SOLENOO-OPERATED VALVES USED IN SAFETY SYSTE MS OF NUCLEAR TUATON ON SAFETY REL ATED EOUIPMENT Evaluston Of Genenc POWER PLANTS Fvaluaten Of Morvionng Methods issue SE NUREG/CR 5700 AGING ASSESSMENT OF RE ACTOR INSTRUMEN-SCIENCE APPLIC ATIONS INTERNATIONAL CORP. (FORMERLY TATION AND POTECTION SYSTEM COMPONENTS Aging Related SCIENCE APPLICATIONS, NUREG/CR-4674 V15: PRECURSORS TO POTENTIAL SEVERE CORE NU G /C 57 9 1 AGING OF NON POWER CYCLE HEAT EL ACCONS W A WM WM Main AM N CHANGERS USED IN NUCLEAR POWER PLANTS Operatmg Empen-NUREG Cf $70 C F SON OF ANALYSIS METHODOLOGIES ^ "

FOR PREDICTING CLEAVAGE ARREST OF A DEEP CRACK IN A RE-ACTOR PRESSURE VESSEL SUBJECTED TO PRESSUR1 ZED THER* NUR CR 5305 VOt: INTEGRATED RISK ASSESSMENT FOR LA-NUREG - 10 AL A O HIGR FUEL RELIABILITY NUREG/CRM59 MODELING fME INF LUENCE OF IRRADIATION NR / 8 R A7HES E. DEPENDENT PROBABILIS-TEMPERATURE AND DISPLACEMENT RATE ON RADi ATION IN' tlc SAFETY ASSESSMENTS Wi1H EMPHASIS ON PRIORITIZATION DUCED HARDENING IN FERRITiC STEEL.s NUREG/CR 5867. GRADIENT STUDY OF A LARGE WELD JOINING NUF EG f673 V03 TR C PF1/ MOD 2 CODE MANUALProgrammer's TWO FORGED A 508 SHELLS OF THE MOLAND PE ACTOR Gade NU E R 5872. ORNOZL: A FINITE ELEMENT MESH GENERATOR i ) Sc 5 FOR NOZZLE-CYllNDE R INTERSECTONS CONT AINING INNER-CORNE R CRACKS SHONKA RESEARCH ASSOCIATES,INC.

NUREG/CR-286 EXPERIMENTAL AND ANALYTICAL INVESTiGATON NUREG/CR4868 DEVELOPMENT OF POSITION SENSfTivE PROPOR-OF THE SHALLOW-FLAW EFFECT IN REACTOR PRESSURE VES- TONAL COUNTERS FOR WJT PARTICt.E DETECT!ON IN LAUNDRY e SELS AND PORTAL MONITORS w NUREG!CR 6003- DENSITY 4 AVE INST ABluTIES IN BO!UNG W ATE R REACTORS TECHNADYNE ENGINEERING CONSULTANTS,'NC.

NURE G/CR-6008 CONSTRAINT EFFECTS ON FRACTURE TOUGH- NUREG/CR 5305 V01 INTEGRATED RISK ASSESSMENT FOR LA.

NESS FOR CIRCUMrERENTIAlt,Y ORIENTED CRACKS IN REAC- SALLE UNIT 2 NUCLEAR POWER PLANT Phenomonology And Rmk TOR PRFSSURE VESSELS Uncertainty Evaluation Prog'am (PRVEP)

International Organization Index This index lists, in alphabetical order, the countries and performing organizations that pre-pared the NUREG/lA reports listed in this compilation, Listed below each country and per-forming organization are the NUREG/lA numbers and titles of their reports. If further infor-mation is needed, refer to the main citation by the NUREG/lA number.

SPAIN MENTED IN 1HE NRC COMPUTER CODES f OR 1HE I4LFLOOD-POLYTf CtiNICAL UMVER3lTY OF MADHID 1940 PHASE OF THE LOCA NUHLG/lA4061. f1EClHCULATION BUCTION LANGE ORLAs, LOCA NUAEG/tA M4 ASSESSMENT OF RELAPS/ MOD 2, CYCLE 36 02, ANALYSi$ OF THE SANT A MARIA DE GARONA NUCL E AR USING NEPTUN REFLOODING EXPER! MENT AL DAT A POWER PLANT USING TRAC.BF1 (G1J1)- NtlRLO/lAuse POS T. TEST. ANALYSIS AND NOOAll2ATION STUDIES OF OLCD LOF1 EXPERIMENT LP-02-6 WITH RELAPS/

BWlTZl.RLAND PAUL SCHtnRER INSTITUTt MOD 2 CY3642-NUREG/tA4x142, DWTFISED FLOW FILM BOLTING AN INVESilGA-TION OF THE POS$1BILITY TO IMPHOVE THE MODELS IMPLE-t 1

43

- - . . . . - . - .= - - -- - . . _ . - , _ . _ - . - -. - - - -. .

t u.a.g. e- at 4 &#-_ -+4e44._- saa4-ama.dJ4.E_ ..*eheaJ-----a.'m--#J-ham *wmJ.J--ee#A,4-.A--ve,.44

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_w- -- .-- erm e L- S -evww9 a e+-m-- e ,r y +ew-r- .r -e --e -- &-- * id r- "- "- '% '*N-e 7 M

• Licensed Facility index This indox lists the facilities that woro the subject of NRC staff or contractor reports. The facility names are arranged in alphabetical ordor. They are procoded by their Docket number and followed by the report number. Il further information is nooded, refer to the main citation by the NUREG number.

3l2'!'? L, "'"

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45

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. -_ -__m. _m .

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t#1C FORM 336 U,8. NUCLEAR f4EGULATORY COMBASSON 1. REPOHi NUMBER (2-69) (Ass 6gned ty t#1C Add Vol.,

P ARCM 1102, Supp., Hev , and Addend.sm Hum.

• * * - "*arI 320b 3* BIBLIOGRAPHIC DATA SHEET (see weiructions on v. fewwse> NUREG-0304

s. in u. Ano suu t a te.

Y"I' I7' N"' 3

3. DAlt HtPOHT PubusMLD Regulatory and Technical Reports (Abstract index journal) uoNm YEAR l

Compilation for November 1992 nird Quarter 1992 ,, ,u On onsyy uvuo,n July-3eptember 6,AUTHOH S; t 6. TYPE OF HEPOH1 Reference

7. PERIOD COVLHED tinclushe Dates)

July-September 1992 8 Pt.HF QHMING OHGAfd/ AisvN = NAML AP maning .dd,.ss; u conn.ctor, p,m.no n.m.O.r.A4)O.HL d m ning .d 68

.s s.pt

PetC. prmede Divisen, Offee er Hagen, U.S. Nucteel Hagulatory Comrruss60n. and Division of Freedom of Information and Publications Services Office of Administration U.S Nuclear Regulatory Commission Washington, DC 20555  !

G OHGAF#Z AIK)N = NAME AND ADOHL89 tif f440, type * $3me as above"; If contractor, prov60s idHC Davision, Office or Hoguan, j 9, SPONbOHN.t

u. a. No. Hegui tory commise nn. nd man,no ede.ss i Same as 8, above. ,

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His journal includes all formal reports in the NUREG series prepared by the NRC staff and contractors, proceed.

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